P. KRAL et al.
Analyses of DEC for NPP’s in the Czech Republic and
their Implementation into SAR
Pavel KRAL
Nuclear Research Institute (UJV) Rez
Husinec-Rez, Czech Republic
Email:
Jelena KRHOUNKOVA
Nuclear Research Institute (UJV) Rez
Husinec-Rez, Czech Republic
Jaromír MACHACEK
CEZ –Temelin NPP
Temelin, Czech Republic
Abstract
The paper is devoted tothe concept of design extension conditions (DEC)and its application and assessment at theCzech nuclear power plants Dukovany (VVER-440) and Temelin (VVER-1000). Starting from evolution of the DEC concept and the role of EUR, WENRA and IAEA – the current status of the DEC concept and its implementation in the Czech Republic is described. The core of the paper is focused on the deterministic safety analyses of design extension conditions without core melt (DEC-A) in the Czech Republic. All major steps and tasks connected with this part of DEC safety assessment are described – methodology basis, role of probabilistic safety assessment, selection of events, computer tools used and their validation, and finally overview and example of safety analysis andincorporation of DEC analyses results into modified format of Safety Analysis Report (SAR).
1.INTRODUCTION
When speaking about safety assessment of design extension conditions, i.e. analyses of events beyond design basis accident, one should distinguish between analyses of DEC without core melt (marked DEC-A in the paper) and analyses of DEC with core melt (marked DEC-B in the paper).
Whereas the later (DEC-B, severe accidents) have been widely assessed and analyzed for at least 2 decades with the accelerator moment of Chernobyl accident, the former (DEC-A, BDBA) were analyzed in the past only partially – typically only the anticipated transient without scram (ATWS) or station blackout (SBO) were analysed and documented in Safety Analysis Report (SAR).
The more systematic work on safety assessment of DEC-A (BDBA) has been started only in the last decade with different starting point and speed in various countries. This effort has been initiated by initiatives and suggestions of European Utility Requirements (EUR)[1], WENRA safety reference levels[7]and IAEA introducing DEC term and concept into the safety standards series[12][14].
The work on BDBA and DEC safety analyses for Czech NPPs was initiated in 2009 as a consequence of the Periodical Safety Review (PSR) after 20 year of the operation of the Dukovany NPP.
2.Evolution of the DEC concept
The term and concept of Design Extension Conditions (DEC) is a new step in evolution of the nuclear safety (see the chronology below). It is a logical follower of previous concept of “design basis” and “design basis accident” supplemented in last decades by analyses of ATWS, SBO and severe accidents (not fully systematic and not reflected in design basis).
Chronology of evolution of basic concept of nuclear safety:
Worst Conceivable Accident (1940’s)
Maximum Credible Accident (1950’s) to
Design Bases Accident, DBA (from 1960’s) to
Plant Design Envelope, PDE incl. Design Extension Conditions, DEC (2010’s)
The term “Design Extension Condition” (DEC) was first introduced in the European Utility Requirements (EUR) in 1992 [1]to define some selected sequences due to multiple failures with the intent to improve the safety of the plant extending the design basis.The DEC are in EUR divided to “complex sequences” and “severe accidents” (corresponding to DEC-A and DEC-B in WENRA terminology).
WENRA published in 2008 document “Reactor Safety Reference Levels” (RL’s)and used firstly the term “design extension” and later in 2014 update of the WENRA RL’s document[7]added clear differentiation between DEC without core melt (DEC-A) and DEC with core melt (DEC-B).
It is worth noting that the evolution of DEC concept was strongly accelerated by the Fukushima accident in 2011 and by following ENSREG activities.
In 2012 the IAEA SSR 2/1[12] introduced the term and concept of“Design Extension Conditions“into the system of IAEA Safety Standards. The concept of DEC was further elaborated in Revision-1 of IAEA SSR-2/1 (2016) and in other IAEA documentsasIAEA SSG-30 (2014), TECDOC-1791 (2016), GSR-Part4 (Rev.1, 2016). DEC will be also an important change in revisions of SSG-2 and GS-G-4.1 guides(to be issued in 2018).Current definition of “design extension conditions” according to Revision 1 of SSR-2/1 (2016) and IAEA Glossary is as follows:
Postulated accident conditions that are not considered for design basis accidents, but that
are considered in the design process for the facility in accordance with best estimate methodology, and for which releases of radioactive material are kept within acceptable limits.
Design extension conditions comprise conditions in events without significant fuel degradation
and conditions in events with core melting.
FIG. 1. Plant states including DEC (according to Rev.1 of SSR-2/1, 2016[12])
UJV Rez has participated actively both in development of EUR (through CEZ membership, from 2007) and in development and revising of IAEA documents.
3.Methodology basis for DEC-A assessment in the Czech Republic
Aside the IAEA recommendations and the Czech Atomic Law, the following regulations, directives and reportsconstitute the legislative and methodological basis for deterministic analyses of DEC-A (BDBA) in the Czech Republic:
—SUJB regulation 195/1999, Requirements on Nuclear Facilities for Ensuring of Nuclear Safety, Radiation Protection and Emergency Preparedness, 1995.
—SUJB directive BN-JB-1.6, Probabilistic Assessment of Safety, 2010(currently revised due to new Atomic Law).
—SUJB directive BN-JB-1.7, Selection and Assessment of Design and Beyond Design Events and Risks for Nuclear Power Plants, 2010[2] (currently revised due to new Atomic Law).
—UJV, Proposal of Methodological Procedure for Performing of Safety Analysis of Beyond Design Basis Accident, UJV Rez, 2010[3].
Analyses of DEC-A scenarios use best estimate computer codes with combination of realistic initial and conservative (or realistic) boundary conditions. The robust design of VVER reactors and their safety features enable to fulfil DBA acceptance criteria in most DEC-A cases including radiological consequences. For the most severe conditions comprising multiple failures of safety systems or safety groups providing protection in the level 3a of Defence in Depth (like SBO), the new measures implemented after post-Fukushima Stress tests in the level 3b of DiD provide additional robust protection against evolution of these scenarios into DEC-B (severe accident).The acceptance criteria applied to DEC-A analyses are identical to those applied to DBA analysis with exception of criterion on primary and secondary pressure and radiological consequences.
The computer code used for NPP safety analyses in the Czech Republic must be approved by the regulatory body according to SUJB directive VDS-030.
4.overall approach to safety assessment of dec in the czech republic
Introduction of DEC concept into area of safety assessment field in the Czech Republic has different impacts in different subareas like probabilistic analysis, deterministic analyses of DBA and BDBA, and deterministic analyses of severe accidents.
Whereas the probabilistic and severe accident analyses were not too much affected by implementation of DEC concept (as the relevant sequences had been analysed before), the deterministic analyses of BDBA (DEC-A) got new strong impulse. And the conceptual and terminological changes in DBA-DEC area are still under evolution.
5.Selection of DEC-A events to be analysed and documented in SAR
The basic set of DEC-A (BDBA) events to be analysed is specified in BN-JB-1.7[2]. Supplemental events and scenarios could be specified by PSA outcomes and engineering judgement.
It is important to mention that in analyses of DEC (which are often complex sequences or combinations of events and failures) it is logical to transfer from “frequency of initial events” to “frequency of occurrence of scenarios”.
SUJB directive BN-JB-1.7 [2]requiresbesides the standard set of ATWS analysesthe following DEC-A (BDBA) events to be analysed:
—Total long-term loss of inner and outer AC power sources;
—Total long-term loss of feed water („feed-and-bleed„ procedure);
—LOCA combined with the loss of ECCS;
—Uncontrolled reactor level drop or loss of circulation in regime with open reactor or during refueling;
—Total loss of the component cooling water system;
—Loss of residual heat removal system;
—Loss of cooling of spent fuel pool;
—Loss of ultimate heat sink (from secondary circuit);
—Uncontrolled boron dilution;
—Multiple steam generator tube rupture;
—Steam generator tube ruptures induced by main steam line break (MSLB);
—Loss of required safety systems in the long term after a design basis accident.
The whole set of prescribed DEC-A analyses was already performed both for Dukovany NPP (VVER-440) and for Temelin NPP (VVER-1000).
As for the documentation of DEC-A analyses in Safety Analysis Report, the temporary solution was creation of a new subchapter 15.9.1 which contains basic results of all DEC-A (BDBA) analyses required by BN-JB-1.7.
The final foreseen solution is introduction of new SAR charter 19, that would contain both DEC-A (BDBA without core melt) and DEC-B (severe accident) analyses presented in systematic and integrated way. Then the Chapter 15 will be again intended for analyses of events up to DBAonly.
Analyses of DEC-A events for the Czech NPP’s have been done with help of RELAP5 computer code. It is worth noting that RELAP5 has been in UJV Rez validated against experimental data from more than 20 tests carried out at various integral test facilities (ITF) and that approximately half of these tests were modelling events of DEC-A type.
6.Example of DEC-A analysis:SBLOCA in VVER-1000 with failure of eccs and operator start of hpsi at 30 min
Analysis of small break loss of coolant accident (SBLOCA) with break D50 mm in cold leg and with failure of start of emergency core cooling systems (ECCS) and operator manual start of high pressure safety injection (HPSI) at 30 min was performed for VVER-1000. Nodalization scheme of VVER-1000 for RELAP5 used and graphical courses of main parameters are shown in Fig.2 – Fig.6 below.
FIG. 2. Nodalization scheme of VVER-1000 for RELAP5 (only primary circuit and 1 of 4 modeled loops depicted)
FIG. 3. Primary and secondary pressure (SBLOCAD50mm in VVER-1000 with failure of ECCS)
FIG. 4. Break outflow and total ECCS injection (SBLOCA D50mm in VVER-1000 with failure of ECCS)
FIG. 5. Reactor levels (SBLOCA D50mm in VVER-1000 with failure of ECCS)
Loss of primary coolant through break D50mm in cold leg and without automatic actuation of ECCS leads to depletion of primary inventory and if not mitigated by operator, core uncovery and overheating. However with respect to high “water volume to power ratio” in VVER-1000, there is sufficient time for operator intervention. In the presented case, operator starts one HPSI pump at 30 min and soon after it the core is quenched and core cooling is restored and stabilized.
FIG. 6. Reactor core temperatures (SBLOCA D50mm in VVER-1000 with failure of ECCS)
7.summary
The work briefly described in the paper has been focused on the implementation of design extension conditions (DEC) concept to safety assessment of Czech nuclear power plants Dukovany (VVER-440) and Temelin (VVER-1000). The core of the paper is focused on the deterministic safety analyses of the design extension conditions without core melt (DEC-A). All major steps and tasks connected with this part of DEC safety assessment are described – methodology basis, selection of events, computer tools used and their validation, and finally overview safety analysesperformed example of results and incorporation of them into modified format of Safety Analysis Report (SAR).
References
[1]EUR, European Utility Requirements for LWR Nuclear Power Plants, Revision D, 2012(the first version A published in 1992).
[2]SUJB directive BN-JB-1.7, Selection and Assessment of Design and Beyond Design Events and Risks for Nuclear Power Plants, 2010.
[3]KRHOUNKOVA, J., KRAL, P., MACEK, J., Proposal of Methodological Procedure for Performing of Safety Analysis of Beyond Design Basis Accident, Revision 1, UJV Rez, 2010.
[4]KRAL, P., Analyses of Beyond Design Basis Scenarios of LOCA with Reduced Availability of ECCS, 2012.
[5]BENCIK M. Analysis of Total Long-Term Loss of Inner and Outer AC Power Sources in NPP Temelin, 2015.
[6]WENRA, Statement and Report on Safety of New NPP Designs, 2013.
[7]WENRA, Reactor Safety Reference Levels for Existing Reactors, 2014(the first version published in 2008).
[8]WENRA, Guidance Document Issue F: Design Extension of Existing Reactors, 2014
[9]SNETP, Identification of Research Areas in Response to the Fukushima Accident, 2013.
[10]NUGENIA, Global Vision, 2015.
[11]OECD NEA, Implementation of Defence in Depth at Nuclear Power Plants, 2016.
[12]IAEA, Safety of Nuclear Power Plants: Design, Specific Safety Requirements, SSR-2/1, Revision 1, 2016 (the first version published in 2012).
[13]IAEA, General Safety Requirements GSR Part 4, Revision 1, 2016.
[14]IAEA-TECDOC-1791, Considerations for the application of the IAEA Safety Requirements on Design, 2016.
[15]IAEA, Safety Classification of Structures, Systems and Components in Nuclear Power Plants, SSG-30, 2014.
[16]IAEA, IAEA Safety Glossary, Terminology used in nuclear safety and radiation protection, 2016.
1