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Project summary


The overall goal of the project is to generate a body of knowledge regarding the uncertainty in the magnitude of fission products release in case of a potential severe accident in Nordic nuclear power plants. The work will provide useful insights into the effect of various types of uncertainty on the source term predictions. Results of the work will be useful for applications that require evaluation of the offsite consequences in terms of public and environmental risks from severe accidents. Data produced in the project can be used for further improvement of modeling in PSA Level 2 and fast source term prediction tools (such as RASTEP).


Assessment of the source terms related to severe accident scenarios is typically performed with integral plant response codes, such as MAAP, MELCOR or ASTEC. These assessments are subject to uncertainty in the accident scenarios (aleatory) and in modeling of phenomena (epistemic).


Typically, the evaluation of the magnitude of the fission products release to the environment is performed for a limited set of accident scenarios, using point-estimate values of epistemic uncertain parameters (user defined sensitivity coefficients in the models, and modelling parameters) in the codes. Furthermore, such analyses typically do not consider the effect of epistemic uncertainty on interactions between physical phenomena and transient accident scenarios, i.e. when physical phenomena (and associated epistemic uncertainty) can significantly affect the course of the accident progression.


For so-called bypass sequences, a standard practice for the sake of conservatism is to define the source term as everything escaping the containment. This creates a situation where a potentially very diverse family of realistic containment bypass scenarios is represented by a set of assumed sequences that contribute substantially to the Large Early Release Frequency (LERF) in a typical PSA L2. In this case, the uncertainty lies in the level of applied conservatism.


In the last few years KTH has developed and demonstrated a systematic approach to quantification of uncertainty in severe accident scenarios and phenomena based on the Risk Oriented Accident Analysis Methodology framework (ROAAM+). The approach combines most recent development in the areas of sensitivity analysis, uncertainty quantification and surrogate modeling approaches. In the previous ROAAM+ work the focus was on the quantification of uncertainty in containment failure probability. The next step in the ROAAM+ development is application to quantification of uncertainty in the source term.


The work is planned to be performed in two phases.


The goal of the first phase of the project will be to identify a set of representative accident scenarios as well as relevant deterministic modelling parameters that can affect accident progression, phenomena and the magnitude of the fission products release.

The selection will be based on the state-of-the-art review of the major contributors to the uncertainty in source term prediction and will include:

Task 1. Review of the safety design of the Nordic BWR.

Task 2. Identification of accident sequences. This will combine information from PSA L1 analysis (plant damage states), potential mitigative\operator actions and timing (EOPs and SAMGs). Furthermore, additional insights from the offsite consequence analysis and emergency preparedness and response will be taken into account during the process of accident sequences definition.

Task 3. Identification of epistemic (phenomenological) modelling uncertainties in different stages of severe accident progression that can affect severe accident progression, release paths and magnitude of the release.

a.         Phenomenology of fission products transport and deposition in the reactor coolant system, steam-lines, turbine and interfacing systems (bypass), containment and reactor building (BWR).

b.         Effect of engineered safety systems on the magnitude of the release.

c.         Review of the modelling approaches and identification of epistemic modelling parameters that can affect the source term in containment bypass sequences.

d.         Identification of potential phenomenological splinters. Phenomena or events where uncertainty is beyond the reach of any reasonably verifiable quantification. These include (but not limited) the mode of the vessel lower head failure, natural depressurization of the primary system due to thermal loading of the main steam lines and\or safety relieve valves, etc.

e.         Mutual interactions between accident scenarios and phenomena (e.g. scrubbing capability of pressure suppression pool in BWRs depends on the temperature of the pool, which is turn depends on accident scenario, including operator actions).


Furthermore, the first phase of the project will include preliminary assessment of the span of possible fission products release magnitudes using MELCOR code. This assessment will be addressed in Task 4.

Task 4.               MELCOR code simulations in the preliminary assessment will be performed for the set of accident scenarios identified in the Task 2, assuming best-estimate and bounding assumptions regarding the values of epistemic uncertain parameters identified in the Task 3. The results of this analysis can be used to screen-out parameters that have negligible impact on the results.


The main outcome of the phase 1 of the project will be a set of accident scenarios that are of interest from both, accident frequency and consequences standpoints, as well as, a set of deterministic modelling parameters that can have a major effect on the magnitude of the fission products release and offsite consequences. Furthermore, it will provide qualitative assessment of the effect of epistemic uncertain parameters on the magnitude of the fission products in different accident scenarios.


The goal of the phase 2 of the project will be evaluation of the sensitivity of the magnitude of the fission products release in different accident scenarios to the variability in deterministic modelling parameters (epistemic uncertainty), identification of the major contributors to the uncertainty, as well as quantification of the uncertainty in the results.

Task 5.               The sensitivity and uncertainty calculations will be performed for the Nordic BWR using MELCOR code and other models and codes dedicated to specific severe accident phenomena. Insights regarding the impact of the results on the analysis of off-site consequences and emergency preparedness and response will be provided. This work will include assessment of available literature as well as relevant new methods concerning source term estimation for containment bypass sequences.  Furthermore, this work will include development and implementation of the algorithms for sensitivity analysis and uncertainty quantification with MELCOR code.


Lead organisation

Vysus Sweden AB


Contact person

Anders Riber Marklund


Contact NKS   NKS Sekretariatet
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DK-4000 Roskilde
  Telephone +45 46 77 40 41

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