Pavlin P. Groudev

Discussion on Practical Elimination of Early or Large Releases for WWER-1000/V320

The paper presents a brief summary of the introduction of the term “practical elimination” as prevention of the conditions that could lead to early or large radioactive releases. The concept of “practical elimination” is defined as part of the Defence in Depth (DiD) of Nuclear Power Plant (NPP) in the International Atomic Energy Agency (IAEA) document INSAG-12 in 1999. But, the special attention to it was paid after the accident in Fukushima NPP in 2011. The mechanisms of the containment failure of reactor WWER-1000/V320 are presented. As an example, the summarized design features and preventing and mitigation measures already implemented at Kozloduy NPP to extend the design basis and beyond design basis envelop are presented. Issues related to external steam explosion are underlined for further study.

Investigation of Quench-16 Experiment With MELCOR Computer Code

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.Copyright

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel.In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed.The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident.The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research.The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.Copyright

Simulation of Quench-12 Test with ASTEC1.3.2 Computer Code

This paper presents an application of the ASTEC V1.3R2 computer code for simulation of QUENCH-12 experiment. The test have been performed to investigate the behavior of VVER type of fuel assemblies during severe accident conditions. In the performed analyses it have been assessed the mass of generated hydrogen during the experiment flooding of overheated core. The base line input model for ASTEC has been provided by Forschungszentrum, Karlsruhe. The comparison of ASTEC1.3R2 calculated results with measured test data shows good agreement.

Verification of Severe Accident Management strategies for VVER 1000 (V320) reactor

Severe accident analysis of a reactor is an important aspect in evaluation of source term. This in turn helps in emergency planning and Severe Accident Management (SAM). The use of the severe accident management guideline (SAMG) is required for accident situation which is not handled adequately through the use of Emergency Operating Procedures (EOP), thus leading to a partial or a total core melt. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). Initiation of SAMG for VVER-1000 is considered at two core exit temperatures viz. 650°C as a desirable entry temperature and 980°C as a backup action [1]. Analyses have been carried out for VVER-1000 (V320) for verification of some of the strategies namely water injection in primary and secondary circuit. These strategies are analysed for a high and low pressure Primary Circuit transients. Station Blackout (SBO) is one such high pressure transient for which core heat can be removed by natural circulation of the primary circuit inventory by maintaining the secondary side inventory. This strategy has been verified where the feed water injection to secondary side of SG is considered from external power sources (e.g, mobile DG sets) as suggested in SAM guidelines. The second transient analysed for verification of the core flooding during Large Loss of Coolant Accident (LOCA) along with SBO, a low pressure event. The injection to secondary circuit is initiated with the available safety pumps and mobile DG sets as soon as feed pumps trip. The analysis shows that SG flooding is not adequate to arrest the degradation of the core. In the second strategy for LOCA transient, the injection to primary circuit has been initiated at 650°C core exit temperature. The analyses show that core flooding is not adequate to arrest the degradation of the core for the large LOCA where as for small break LOCA the injections through available safety systems are adequate. The assessments are carried out with integral severe accident computer code ASTEC V1.3.

Results of the QUENCH-L2, DISCO-L2, and COMET-L2 Experiments Performed Within the LACOMERA Project at the Forschungszentrum Karlsruhe

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: Fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.Copyright

RELAP5/MOD3.2 Investigation of a VVER-1000 MCP Switching on Problem

This paper provides a discussion of various RELAP5 parameters calculated for the investigation of the nuclear power reactor parameter behavior in case of switching on one main coolant pump (MCP) when the other three MCPs are in operation. The reference power plant for this analysis is Unit 6 at the Kozloduy Nuclear Power Plant (NPP) site. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. This investigation has been conducted by Bulgarian and Russian specialists on the stage when the reactor power was at 75% of the nominal level. The purpose of the experiment was the complete testing of reliability of all power plant equipment, testing the reliability of the main regulators and defining a jump of the neutron reactor power in case of switching on of one main coolant pump. The Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and Kozloduy NPP have been developing a RELAP5/MOD3.2 model for Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.Copyright