[en] Computer codes are being used to analyse operational events in nuclear power plants but until now no special attention has been given to the dissemination of the benefits from these analyses. The IAEA's Incident Reporting System contains more than 3000 reported operational events. Even though deterministic analyses were certainly performed for some of them, only a few reports are supplemented by the results of the computer code analysis. From 23-26 May 2005 a Technical Meeting on Deterministic Analysis of Operational Events in Nuclear Power Plants was organized by the IAEA and held at the International Centre of Croatian Universities in Dubrovnik, Croatia. The objective of the meeting was to provide an international forum for presentations and discussions on how deterministic analysis can be utilized for the evaluation of operational events at nuclear power plants in addition to the traditional root cause evaluation methods

[en] The paper presents the results of the independent analysis of the operational event which took place on 07.11.2003 at Unit 1 of Rostov NPP. The event started with switching off the electrical generator of the turbine due to a short cut at the local switching substation. The turbine isolating valves closed to prevent damage of the turbine. The condenser dump valves (BRU-K) and the atmospheric dump valves (BRU-A) opened to release the vapour generated in the steam generators. After the pressure decrease in the steam generators BRU-K and BRU-A closed but one valve stuck opened. The emergency core cooling system was activated automatically. The main circulation pump of the loop corresponding to the steam generator with the stuck BRU-A was tripped. The stuck valve was closed by the operational stuff manually. No safety limits were violated. The analysis of the event was carried out using ATHLET code. A reasonable agreement was achieved between the calculated and measured values. (author)

[en] In relation with a real operational event, when due to some unexpected and unnoticed conditions, strong temperature stratification developed in 3 of the six loops of the unit 1 of Paks NPP. During the incident, which lasted for about 8 hours, in several cases the potentially critical stresses may have developed in the cold legs and also in the reactor vessel. It was extremely important to evaluate the potential consequences of the incident. However, it turned out that that the both the available set of measured information, both the available analysis tools were not completely satisfactory to reveal all the details of the operational sequence. The paper outlines the methodology applied to extract the highest possible information from the available measurements, and also the efforts the set up and apply analytical models do simulate those physical parameters which could not be extracted from the measured information. At the end, the mail lessons to learn from the incident are summarized. (author)

[en] Based on review of cooperation of the VUJE, Inc. with Slovak NPP operator within the last three years, the paper describes selected operational occurrences, which were required to be analyzed from thermohydraulic point of view. For each event a short problem description is given, followed by information of analytical methodology and approach, as well as overview of the most important findings. The events, described in detail include problems with mechanical fatigue of the pressurizer surge line due to the thermal stratification in the upper part of the line (CFD simulation using FLUENT code), leakage from the primary circuit due to the thermal stress and fatigue of the ECCS pipelines (RELAP5 simulation of the ECCS line section with two back valves) and low-level leakage through the reactor flange sealing (RELAP5 simulation together with activity analysis). (author)

[en] At present time a problem of obtaining from NPPs the precise information on operational events in reactor facility, which used then for performing the safety analysis and validation of input models is very topical. The comparative analysis of data from data-processing system (DPS) acquired upon operational events on reactor facilities VVER-1000 (V-320) of KhNPP unit 1 and 2 is cited in given report. Tripping of one and two feed-water pumps (FWP) were selected as operational events. Data of KhNPP unit 1 were acquired from DPS of outmoded type (output of eighties). The problem of KhNPP unit 1 data is in large discontinuity of derived data upon operational events and it has a great effect on accuracy of calculations. Data of KhNPP unit 2 were acquired from modern DPS, which was installed and modernized directly before unit's setting in operation in 2002-2004. In this case we have practically continuous array of data on operational event. Since the first fuel load on KhNPP unit 2 improved fuel assemblies (TVSA) are used, some of which have an embedded temperature detectors. Improved construction of fuel assemblies allow to define the temperature fields and distribution of energy-release in active core more exactly. The data of real operational events were also compared with data acquired during the calculation of presented operational events with use of thermohydraulic code RELAP5. The results of given analysis are used for input models of presented reactor facilities validation and allow to conclude that model of KhNPP unit 2 is performing the analysis of different operational events more precisely. It is possible to make a conclusion as a result of performed comparative analysis to the effect that accuracy and quantity of data are making a great effect on accuracy of further safety analysis of corresponding NPP units and also about the necessity of modernizing of outmoded DPSs for the purpose of increase of calculation accuracy hence for increase of unit's safety in whole. (author)

[en] Among the various efforts to improve operational safety of nuclear installations, systematic collection, evaluation and feedback of operational experience are considered valuable and effective. Such a system enables all safety related events to be analyzed for determination of the root causes and necessary corrective and preventive action to be taken to avoid their recurrence and to enhance operational safety at Nuclear Power Plants. The goal of event investigation is to improve overall plant safety and reliability of operations by learning from experience. Although use of probabilistic methodologies is gaining pace, the regulatory authority in Pakistan (PNRA) mainly relies on deterministic analysis for operational safety at NPPs. This covers the determination of safety margins and root causes of events and the necessary evaluations for verification of results of assessments conducted by the NPPs. This paper provides a general overview of events reported by NPP licensees to PNRA and includes examples of evaluations carried out by PNRA with the regulatory perspective. Among the various evaluations carried out by PNRA, one such evaluation of operational events included in this paper as an example relates to control rod drop time of our 300 MW(e) NPP at Chasma in Pakistan. In this event the drop time, during the test, was found to be higher than the required drop time i.e. <2 seconds for two control clusters. For this case, evaluation of the neutronic analysis for most significant design bases condition was conducted by PNRA to assess the safety margins. In addition, PNRA also conducted analysis of the thermal hydraulic characteristics in the core and possible flow induced effects on mechanical components which included determining the effects of transverse flow forces and fluid densities at various temperatures on slow drop movement of the rods. (author)

[en] Deterministic approach requires accurate evaluation of uncertainties. The uncertainties determined as result of computer code calculations have different structure than the uncertainties resulting of parameters measurement on real unit. Understanding of these values is an important issue for validation of deterministic computer code input data. Therefore a special attention must be given to collection and analysis of operational parameters. Since 1996, complex TOFT has been consequently installed on units equipped with WWER-440 reactors of 'Kozloduy' NPP. One of the applied approaches for analysis of the accumulated archives, is pure statistical. It is based on methodology adopted from mathematics of communications, where analysis of transmitted electric signals is done in order to get higher level of measurement accuracy. The main considerations of the methodology are described in the paper. It includes description of a way for evaluation of the central moments in form of location, scale and kurtosis. Shannon's entropy, calculated in form of entropy coefficient is evaluated for the purpose of identification the possible probability density function and the ranges of particular parameter uncertainty. As result there are produced cumulative uncertainties of signal transmission, which include: perturbations in measured medium, sensor properties, intermediate signal transformations and the features of output signal. Identification of the uncertainties of important measured parameters during the unit normal operation is performed. There is presented statistical analysis of core inlet/outlet temperatures, pressure of primary circuit and etc. on different operation states, followed by the corresponding conclusions. Six basic states are considered in the analysis. In addition, evaluation of core channels outlet temperatures is done, where outlet thermocouples are analyzed according to the presented methodology. (author)

[en] In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIVs) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed assuming realistic equipment behavior and operator actions. The comparison of the RELAP5/MOD 3.3 results for the realistic analysis with the measurement has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure) for both power reduction transient (100 - 28 %) and pump trip event. Four additional RELAP5/MOD3.3 analyses with different transient scenarios that contribute to Conditional Core Damage Probability (CCDP) in the pump trip event were performed. (author)

[en] For the cooling of water forced circulation through the RBMK-1500 reactor at the Ignalina Nuclear Power Plant eight Main Circulation Pumps (MCPs) are employed. These pumps are joined into groups of four pumps each (three for normal operation and one on standby). There were few events when one or few MCPs were inadvertently tripped. On 14 May, 1996 one MCP at Ignalina Unit 2 was inadvertently tripped. The similar event took place on 23 January, 1998. During this event, the MCP check valve failed to close, causing a recirculation loop to develop by means of a reversed flow through tripped pump. On 31 July, 2000 three MCPs at Ignalina Unit 2 were tripped one after another, due to inadvertent activation of fire protection system. Simultaneous trip of all MCPs occurred on 26 March, 1986. In the case of one MCP trip the throughput of the rest running pumps in the affected Main Circulation Circuit loop increased, however, the total coolant flow through the affected loop decreased. The main question arises whether this coolant flow rate is sufficient for adequate core cooling. In the case of all MCPs trip, the coolant during the first few seconds is supplied to the reactor by pumps coastdown. Later the reactor is cooled by natural circulation. This paper presents investigation of one and all MCP trip events at the Ignalina NPP by employing best estimate code RELAP5. For single tripped MCP and simultaneous trip of all MCPs cases uncertainty and sensitivity analysis was performed. For that purpose the GRS (Germany) developed System for Uncertainty and Sensitivity Analysis package was used. Uncertainty analysis of flow energy loss in different parts of the Main Circulation Circuit, initial conditions and code-selected models was performed. On the basis of these analyses recommendations for the improvement of the Ignalina NPP RELAP5 model have been developed. (author)

[en] An adequate level of safety is achieved in nuclear power plants by a suitable design of protection systems. Safety Analysis Reports typically include a verification that the plant behavior under a set of Design Basis Accidents (DBAs) (including from anticipated operational events to postulated accidents) remains under specified acceptable limits and, therefore, enough safety margin is provided by the protection design. Conclusions of the Design Basis Accident Analysis (DBAA) depend on a set of assumptions on initial and boundary conditions and on system reliability which should be consistent with the operation of the plant. Many of these assumptions are imposed as Technical Specifications or equivalent operation requirements. This kind of safety verification, however, is mainly analytical and only very specific aspects of the protection design can be experimentally verified. Operational events are unique occasions to check, on one hand, if the plant protection behaves as expected and, on the other, if the designed protection is enough to guarantee a sufficient level of safety. There are several mechanisms of safety margin degradation. Some are of dynamic nature, i.e., the plant or protection behavior is not as expected and a loss of margin may occur, eventually leading to limit exceedance. Others are of probabilistic nature, for example, due to a loss of protection reliability leading to a given probability of limit exceedance. Incident analysis should address, to the extent possible, all these mechanisms. As a part of the analysis of operational events it should be identified whether the incident meets the assumptions of the DBA analysis or it is outside the design basis envelope. In the former case, one or more DBAs can be identified as envelope transients. Thus, safety margins during the real event can be verified to be equal to or greater than those demonstrated in the Design Basis Accident Analysis for the envelope transients. In the latter case, it should be identified if one or more of the safety limits used as acceptance criteria for DBAs has been actually exceeded. In any case, an extension of the precursor analysis techniques, usually focused on the eventuality of degradation to severe accident conditions, can be used as a means to evaluate probabilistic mechanisms of safety margin loss. (author)