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Need and Alternatives for Filtered Venting of the Containment

Need and Alternatives for Filtered Venting of the Containment


Beneficiary Organization Details:

Paks NPP (Hungary):

Project Aims

The main purpose of this project was to develop safety goals and functional requirements for filtered venting system, or alternatives thereof, to evaluate potential system for VVER 400.213 application, and provide recommendations for the selection of a system that satisfies safety goals and meets the functional requirements.

The analyses were performed by VEIKI Budapest, Hungary and NRI Rez, Czech Republic, using the MAAP4/VVER code with the parameter file for Paks NPP. The objectives of the projects were to (1) Assess the possibility to install a filtered vent system (FVS) to limit containment pressure rise and release to the environment during a severe accident, and (2) Define the characteristics of such a FVS. Based on these objectives, individual project tasks were laid down:

  • Develop design bases and functional requirements;
  • Evaluate existing FVS for VVER 440/213 applications;
  • Analysis and description of proposed FVS;
  • Develop pressure/release control severe accident guidelines;
  • Final report and recommendations.

Westinghouse Electric Europe s.p.r.l. (WEE) took the lead as main contractor for the projects, with subcontractors VEKI Budapest (Hungary), NRI Rez (Czech Republic) and NSS Ganserndorf (Austria) forming the whole project team.

In order to meet safety goals and requirements, the design basis established for this projects were the following; (I) the primary design basis that the fission product release form from an intact containment should not exceed 0.1% of Cs inventory in any case, (II) the secondary design basis concerning containment load, requiring that the absolute pressure 0.3 MPa should not be exceeded during long term pressurization.

The following sever accident sequences were studied (PH2.06/94):

  • Loss of all AC power with operator action to depressurize the system at core damage;
  • Small break LOCA (ID 25mm) with failure of safety injection;
  • Large LOCA (double ended, ID 500mm) with failure of safety injection.

Analyses were performed exclusively with the Modular Accident Analysis Program, VVER version (MAAP4/VVER). This code performs an integrated simulation of accident conditions in VVER type reactors, including severe accidents involving core melt, vessel failure and ex-vessel thermal hydraulics, fission product and structural response. The code is well suited to application such as feasibility and design studies for severe accident mitigation measures, and accident management investigations. Code modification addresses the inclusion of hydrogen recombiner models and inserted gas injection; in addition a decoupled code version was developed to separate the containment from the primary system. The code underwent an extensive qualification program within the framework of earlier Phare 4.2.7.a project that demonstrated that MAAP4/VVER is a suitable tool for performing required analyses.

Plant design data were taken from the Paks NPP (Hungary), which represents a standard plant with Russian designed VVER 440/213 reactors.

Project Results

The results of the project are considered decisive for limiting the consequence of severe accidents in VVER 440/213 reactors. Specific design features of this plant strongly influence safety goals for filtered venting system and optimum hydrogen control strategies. In contracts to typical Western PWRs, due to significantly higher leakage rate of an even intact containment, the primary safety goal of filtered venting system is the limitations of radioactive releases, and prevention of containment overpressurization being considered the secondary safety goal. Severe accident hydrogen is identified as the major threat to containment integrity and needs to be dealt with by suitable mitigation devices.

An important step of the project was the quantification of safety goals relevant to filtered venting system and optimum hydrogen control strategies for the specific type of plant. These quantified safety goals were derived during the project meeting and agreed upon by the Nuclear Safety Authorities of the beneficiary countries as a basis for this project.

Comprehensive analyses of a large number of LOCA with various break sizes as transients were performed to study important severe accident phenomena and investigate possible solutions to limit radioactive releases, prevent late containment failure and cope with sever accident hydrogen. Heavy reliance was placed on results of analyses performed under PHARE 4.2.7.a project on beyond design basis accident analysis and management, and full of those results were made.

The conclusions of the study led to recommendations on the means to be implemented to satisfy safety goals. Safety criteria and design bases were developed based on technically sound arguments and are consistent with those adopted in various foreign countries. It is technically challenging to meet the targets chosen, but has been shown feasible. It is important to note that these safety criteria /system design bases have a potentially large impact on the final recommendation on system upgrades. For example, if decision were made to modify the release target or the lower limit for deflagration to detonation transition, the requirements for system upgrade would change considerably.

There is interaction between measure proposed for limiting radioactive releases and control hydrogen during severe accidents. To address interactions between all measures proposed for the mitigation of severe accident phenomena, a plant specific Level 2 PSA is the most appropriate tool. This would also assist in decision-making regarding sever accident mitigation system and guideline implementation.

Radioactivity release control during accident in-vessel phase:

Installations of a severe accident backup spray system, with nominal flowrate of 2*600 m3/h, actuated by operator action when the core exit temperature reaches 6500C. The water source available should be sufficient to allow long-term operation of the system beyond the end of release time. This can be achieved using a combination of dedicated water supply, and connections to existing large capacity uncoated water sources. The recommendation does not involve filtered venting during the early (in-vessel) stage of the accident, nor does it require that once initiated, the spray is controlled. However, reactor system depressurization and prevention of long term overpressure are important accident management measures, which also contribute to release control.

Control of long term pressure rise in the containment:

Long term venting of the containment may not be required depending on concrete composition and the natural containment leakage rate. If it is, it should be achieved using the existing TL 70 ventilation system. Suitable operator guidance can be included in the Severe Accident Management Guideline (SAMG) package.

Use of filtered venting system to meet primary design basis for most severe accident would be very difficult, the system would be large and complex, and the cost of such system would be excessive.

Further Project Results

Further information on the project results could be sought from the beneficiary organization.

General Information

Need and Alternatives for Filtered Venting of the Containment
Budget year: 
Types of activities: 
Design safety
Installation types: 
VVER 440-213
Duration (months): 
Electric Power Research Institute VEIKI, Hungary; Nuclear research institute Rez, Czech Republic; Nuclear Safety Services, Austria
Old reference: 
Effective contract date: 
Closure date: