Competence fields
The activities of NUBIKI serve to ensure safe operation of nuclear
power plants. NUBIKI staff members perform deterministic, as well as
probabilistic analyses to improve the safety level of nuclear power
plants.
In the frame of deterministic
safety analyses, NUBIKI performs computational simulations to
analyse different event sequences which can occur in a nuclear power
plant. One of the fields of our activities is the analysis of the
containment response in case of design basis accidents using computer
codes. The computed results are validated with post-test calculations
of experiments. Important condition for the safe operation of a plant
is the containment leak rate analysis. Our traditional field is the
compilation and evaluation of the measured data of numerous leak test
measurements performed since the commissioning of the power plant.
Thermal-hydraulic processes and activity release during severe accident
sequences has been analysed applying computer models. In this field
analyses in the AGNES project dealing with the safety analysis of the
Paks Nuclear Power Plant, and during the project for the Level 2
probabilistic safety assessment of the plant have been done. Important
field of the severe accident analyses is the evaluation of the hydrogen
distribution and management in the containment. For this purpose CFD
tools are used. Our computer simulations – among others – support the
development of safety management guidelines of the plant. NUBIKI
participates in several international experimental projects with
post-test calculation of experiments. The objective of this work is the
verification and validation of computational models.
We utilise probabilistic safety
analyses to identify event sequences that lead to a severe accident
and to determine the likelihood of such sequences. We develop detailed
event sequence models for a wide range of accident scenarios. The
responses of plant systems, system components and the interactions of
plant personnel are all subject to in-depth analysis. We apply
state-of-the-art analysis methods and computerised tools for model
development and quantification. Over and above a quantitative
description of nuclear safety, use can be made of our analysis results
in support of risk-informed decision-making. The nuclear safety
authorities as well as the utilities can take advantage of this
decision-making process. We devote substantial efforts to developing
appropriate methods and tools for risk-informed applications.
Nowadays we make uses of our experience and expertise from more than
two decades of probabilistic safety assessment in providing support to
plant time extension activities, design and implementation of plant
changes, making decisions related to plant operation and maintenance
and preparation for new nuclear builds in Hungary.
We regularly apply deterministic and probabilistic analyses in
combination as complementary methods.
Deterministic safety analyses
Containment studies
The team has a long-term history in performing DBA calculations for
the VVER-440 containment. The applications included the update of DBA
analyses for the Final Safety Analysis Report (FSAR) of Paks NPP.
Recently uncertainty analyses were performed to explore the margins of
the containment parameters under DBA conditions.
Severe accident analysis
The behaviour of the VVER-440 plant under severe accident conditions
was studied with code calculations in the AGNES project, and later in
several PHARE projects.
Further studies involved accident progression and fission product
transport analyses in the Level 2 PSA project for Paks NPP. Level 2
analyses also served as a basis for the development for a severe
accident management strategy for the VVER-440 plant.
Severe accident management
Verification analyses of the severe accident management guidelines
(SAMG) with code calculations were performed to support the development
of the guidelines and to check the time frames available for the
interventions. Hydrogen distribution in the containment and hydrogen
management with passive catalytic recombiners was analysed with 3D
containment calculations.
Probabilistic safety analyses (PSA)
We apply level 1 probabilistic
safety assessment to analyse and evaluate accidents that may result in
severe damage in the reactor core or in the fuel assemblies stored in
close vicinity to the reactor. We have completed such analyses for a
wide range of potential initiating events and operational states of the
four Paks nuclear power plant units. We have studied internal events
and failures (e.g. pipe ruptures, system malfunctions), internal
hazards (on-site fires, internal flooding) as well as external events
(earthquakes) as initiating events that divert the plant from normal
operationally conditions. As to plant operational states, our analyses
cover plant operation at full power and low power and shutdown states
of the reactor too. The scope of our PSA for fuel damage accidents in
the spent fuel storage pool within the reactor hall of the Paks plant
is comparable to that of the reactor PSA. We apply internationally
recognised computerised tools in support of model development, input
data compilation and analysis, and risk quantification. Moreover, we
also make use of specific analysis tools that are the result of our
in-house developments.
We develop level 1 PSA further up
to level 2 probabilistic safety assessment. We determine the potential
large releases of radioactivity to the environment and the likelihood
of such consequences by taking into account the severe accident
processes within the containment. Similarly to the level 1 analyses,
the level 2 PSA for the Paks nuclear power plant include a broad
spectrum of potential severe accidents. The Risk Assessment Division
and the Safety Analyses Division of our institute work in close
co-operation to perform level 2 PSA.
We regularly review and update
the existing probabilistic safety assessment studies for the Paks
nuclear power plant including PSA model, reliability data, analysis
results and documentation. Usually an update is made every year to
ensure a credible picture about the safety level of the plant. During
updates we give considerations to safety related plant changes, new
experiences from plant operation and maintenance, extensions to
analysis scope, and application of advanced analysis methods and new
knowledge.
In our efforts towards method
development we put much distinguished emphasis on the establishment of
models and approaches that enable a probabilistic description of
interactions made by the plant personnel. We have performed numerous
experimental studies, observations at the full scope, replica training
simulator of the Paks plant. We collected data on the characteristics
of control room crew operations in emergency during the observations.
Simulator data were subsequently used in combination with expert
opinion in direct support to method development. We have made uses of
these methods in human reliability analysis as part of the plant PSA.
From among the analysis tools we
developed the so-called ADRIA computer code should be highlighted.
ADRIA is a combined database, analysis as well as design tool we
constructed originally for the purposes of fire and flood PSA. ADRIA
includes a detailed inventory of plant components including cables and
cable routing in particular. In addition to its uses in risk
assessment, it has also important functions in designing cable routes
and providing support to operational tasks and decisions related to
cabling. In order to enable lifetime extension of the Paks plant ADRIA
is applied in the ageing management program for cables. Also, it can be
used to underpin decommissioning activities when the plant has come to
the end of its lifetime.
Over and above the fulfilment of certain nuclear safety
requirements, probabilistic safety assessment is utilized in actual
PSA-applications within a risk-informed decision-making framework.
Activities of both the nuclear safety authority and the licencsee can
be supported by these applications.
If
justified by safety analysis results, we develop proposals for plant
changes to increase safety by removing vulnerabilities to accidents. We
determine the expected improvement in plant safety if such changes are
made. We have performed risk assessment in the design and in the
implementation phases of several safety related modifications made to
the Paks nuclear power plant. These assessments were of high importance
during a large-scale safety improvement program that was implemented
between 1996 and 2002 at Paks.
We
developed a proposal for the Hungarian Atomic Energy Authority to
develop and implement risk-based safety indicators that could be used
to monitor and predict the safety performance of a nuclear power plant.
As one element of the proposed indicator system we have adapted
analysis and evaluation method used in the US to determine the plant
level risk impact of events occurred at nuclear power plants. In order
to make such analyses simpler we have created a computer programme that
uses the level 1 PSA model of the Paks nuclear power plant. We make use
of this programme and the adapted analysis method during the analysis
of events reportable to the nuclear safety authority by the Paks plant.
We determine the conditional core damage risk for each and every
licensee event as a regular support to the authority.
We have
developed a computer code that is capable of calculating core damage
risk and displaying risk graphically in accordance with the actual
configuration of safety related systems and components in a nuclear
power plant. This programme is called Risk Supervisor and it is
intended for regulatory use. Similarly to PSA-based event analysis, we
analyse all the reported licensee events by the Risk Supervisor too.
We also develop a risk monitor for the four units of the Paks nuclear
power plant. This risk monitor builds upon the RiskWatcher risk monitor
software of Scandpower AB and the unit specific PSA models. It can
calculate and evaluate configuration specific risk in a more detailed
manner than the Risk Supervisor and it can also take into account more
factors that could influence risk. It is envisaged that the risk
monitor will be the basis of risk-informed decision-making in relation
to a number of operational and maintenance tasks.
We have
constructed a computerised tool that can be used to determine the
likelihood of a severe accident during plant emergency. This tool is
installed at the Centre for Emergency Response and Training Awareness
of the Hungarian Atomic Energy Authority. If an event calls for the
emergency operation of the centre, then this computer program can
calculate the conditional probability for the development of a core
damage accident. The precision of the prediction is driven by the scope
and level of detail of the information available about the actual plant
event.
We are
engaged in method development to enable the use of risk information
(derived from probabilistic safety assessment) besides the traditional,
deterministic safety classification process for the systems and
components of a nuclear power plant. At an early stage of thus
development we determined the safety significance of systems and
components modelled in the plant PSA. We seek for combined, mutually
complementary uses of deterministic principles and quantitative risk
information in a common framework. The key objective of this effort is
to enable the uses of safety requirements that differentiate in
accordance with an improved, more risk-informed safety classification.
We have
reviewed and evaluated the potential areas of risk-informed
decision-making that can be supported by probabilistic safety
assessment. We have produced a series of technical reports in this
subject within our consultants’ services to the Paks Nuclear Power
Plant Ltd. and, also, as a technical support organisation to the
Hungarian Atomic Energy Authority. In these reports we identified the
potential applications the underlying objectives, as well as the
technical and administrative conditions for their introduction. Use is
to be made of these reports in the current co-ordinated efforts of the
authority and the licensee towards risk-informed decision making in
Hungary.
We have
developed methods for risk-based review and modification of the
surveillance test intervals and the allowed outage times of safety
related systems and components laid down in the Technical
Specifications of the Paks Nuclear Power Plant Ltd. We have made a
number of trial applications to demonstrate the use of our proposed
method for revising allowed outage times. As to surveillance test
intervals, we have programmed our methods in a computer code and we
have applied it the full range plant systems modelled in the PSA for
the Paks plant.
We have
studied and evaluated to what extent the training scenarios included in
the continuing training programme for the control room crews of the
Paks nuclear power plant cover the accident sequences found dominant in
the plant PSA.
We utilise the methodology of probabilistic safety assessment and
also the experience from such studies indirectly in support of
strategic tasks aimed at ensuring long-term safe operation of the Paks
nuclear power plant.
The
summary description of probabilistic safety assessment is part of the
Final Safety Analysis Report which is the most important document that
demonstrates the safe operation of the Paks nuclear power plant by
describing and evaluating the compliance with nuclear safety
requirements. Moreover, we have performed fault tree based reliability
analysis for a range of plant systems in order to enable an evaluation
of the numerous nuclear safety requirements related to system
reliability. Theses system level analyses form the basis of safety
assessments presented separately for the different systems within the
Final Safety Analysis Report for the Paks plant.
We have
elaborated a guide on the methodology to monitor the effectiveness of
maintenance by regularly keeping track of and evaluating performance
indicators for safety related plant systems at the Paks nuclear power
plant. We have identified the performance indicators for all the plant
systems that are subject to monitoring and we have determined the
acceptance levels of these indicators based on fault tree analysis.
Also, we provide support to the plant in implementing the proposed
monitoring programme which is a necessary condition for lifetime
extension at Paks.
We utilise our PSA knowledge and
skills to contribute to the ongoing scientific and technical
preparatory work aimed at extending the capacity of nuclear power at
Paks by building new reactor units. Further details on this subject can
be found under References
within this website.
International co-operation plays
an important role in our analyses and also in the underlying method
developments. An example of such co-operation is the risk assessment of
the Proton Therapy Facility erected for cancer treatment at the Paul
Sherrer Institute in Switzerland. We have contributed to that
assessment by developing fault tree models for the electrical and
control and instrumentation systems of the facility.
For further information on probabilistic
safety assessment and its applications please contact Attila Bareith,
Head of Risk Assessment Division, by e-mail or phone (+36 1 392 2716).
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