FORESEEN IN THE FRAMEWORK OF PLANT LIFE EXTENSION
R. COZERET IRSN
Fontenay-aux-Roses, France Email: romain.cozeret@irsn.fr
C. DEBAUDRINGHIEN, G. CENERINO, E. RAIMOND IRSN
Fontenay-aux-Roses, France Abstract
In France, EDF is developing a Plant Life Extension (PLE) program for the Gen II PWRs, which both takes into account the lessons of the Fukushima Dai-chi accident and aims at reducing the gaps in terms of safety with the Gen III EPR, as requested by the French Safety Authority ASN. This program is progressively reviewed by IRSN for the ASN. The paper presents some intermediate statements of this review for the upgraded strategies proposed by EDF in order to reduce the consequences of a severe accident on a Gen II PWR. It gives some comparisons with the Flamanville 3 EPR.
1. INTRODUCTION
The French electrical utility EDF is currently operating a fleet of 58 Gen II Pressurized Water Reactors (PWRs) (900, 1300 and 1450 MWe series) built between 1977 and 1999. Periodic Safety Reviews (PSRs) are conducted every 10 years. These reactors were not designed to face a core melt accident and several plants reinforcements have been discussed in France and progressively implemented by EDF to allow for the management of severe accidents.
In 2009, EDF presented to the French Safety Authority (ASN) a Plant Life Extension (PLE) program, in order to give a possibility to extent the Gen II PWRs operation duration beyond 40 years. It included an ageing program but also some reinforcements to reduce the gap with the safety objectives of the new nuclear power plants like the Gen III EPR. This program has been reviewed by IRSN for ASN in parallel with the Complementary Safety Assessments (CSA) after the Fukushima Dai-ichi accident.
Today, the PLE and the CSA programs are combined in a vast industrial project by EDF with also some important efforts from IRSN and ASN to review the different steps of the project until final implementation.
The paper summarizes some of IRSN statements after the review of the EDF upgraded severe accident mitigation strategies for the generation II PWRs (firstly 900 MWe series) foreseen in France in the frame of PLE and CSA. This review has been presented in July 2016 to the French Advisory Committee for Nuclear Reactor Safety and ASN.
2. FRENCH GEN II PWRS BACKFITTING FOR SEVERE ACCIDENT
Before presenting the future PWRs safety reinforcements for severe accident, it is important to remind some NPPs upgrades that are or are being implemented on the French Gen II reactors [1]:
— Development and update of severe accident management guidelines;
— Installation of an Emergency Filtered Containment Venting System (EFCVS), with some specific procedures which can differ from one reactor to the other;
— Installation of Passive Autocatalytic Recombiners (PARs);
— Reinforcement of the closure system of material access penetration for the 900 MWe PWRs reactor building (above the design pressure);
— Reactor cavity reinforcement (basemat width and corium spreading area) for Fessenheim NPPs;
— Instrumentation to detect hydrogen in the reactor containment;
— Instrumentation to detect a vessel failure;
— Modification of the pressurizer safety valves (reliability in case of station black out);
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— Reinforcement of electrical supply of the containment isolation system and optimization of procedures for the manual actions;
— Reinforcement of the ventilation system of the secondary reactor building for the 1300 and 1450 MWe PWRs;
— Re-injection of contaminated water from auxiliary buildings (in case of leakage on sump recirculation circuits) to the reactor building (for 1300 MWe PWRs).
All these backfittings have been decided and designed based on the knowledge obtained from research on severe accident and learnings obtained from deterministic safety studies or Level 2 Probabilistic Safety Assessment (L2 PSA) studies.
In parallel, EDF is conducting a verification that the equipment and structures which are needed to mitigate a core melt accident can withstand to the severe accident conditions. If any, the limits to the equipment robustness have to be known.
Even if all these reinforcements bring very substantial risk reduction, especially for the short term of any severe accident, it has to be recognized that there are some important gaps with the solutions developed for a Gen III reactor like EPR.
For example, in the worst severe accident situations (typically a long term total loss of core cooling), it is difficult to demonstrate that the corium can be stabilized after vessel failure. Another important limit is the protection of severe accident equipment against external hazards (earthquake, especially) which was not considered before the Fukushima Dai-chi accident. That means that the efficiency of the long term accident management strategies for the Gen II PWRs is still limited in comparison with the efficiency of the EPR strategies.
For IRSN, these limitations, and also the progress in the knowledge on severe accident progression, justify to continue efforts in Gen II PWRs reinforcement for severe accident.
3. GEN II PWRS SAFETY OBJECTIVES ASSOCIATED TO THE REDUCTION OF POTENTIAL RADIOACTIVE RELEASES IN THE FRAMEWORK OF PLE
In the framework of the Gen II PWR program, the French Nuclear Safety Authority stated to EDF that the safety objectives of the Gen III reactors (for instance the Flamanville 3 EPR) should be used as a reference for all studies undertaken in the frame of PLE. For EPR, the general objective [1] related to the reduction of potential radioactive releases is“to achieve a significant reduction of potential radioactive releases due to all conceivable accidents, including core melt accidents”. It means:
— “For accident situations without core melt, there shall be no necessity of protective measures for people living in the vicinity of the damaged plant (no evacuation, no sheltering);
— Accident situations with core melt which would lead to large early releases have to be "practically eliminated" : if they cannot be considered as physically impossible, design provisions have to be taken to design them out. This objective applies notably to high pressure core melt sequences;
— Low pressure core melt sequences have to be dealt with so that the associated maximum conceivable releases would necessitate only very limited protective measures in area and in time for the public. This would be expressed by no permanent relocation, no need for emergency evacuation outside the immediate vicinity of the plant, limited sheltering, no long term restrictions in consumption of food”.
The first objective is mainly related to Steam Generator Tube Rupture (SGTR) and Loss of Coolant Accident (LOCA) design basis accidents. The EDF proposals for the 900 MWe PWRs PLE program have not been reviewed yet by IRSN and are not developed in this paper.
The second objective has been largely addressed by the modifications which have been summarized above.
For the third objective, two remaining issues have to be considered:
— The long term stabilization of the corium in case of vessel failure cannot be demonstrated for all conditions;
— The EFCVS decontamination factor for gaseous iodine is not sufficient to limit the need for emergency evacuation to the immediate vicinity of the plants.
EDF, in accordance with its initial PLE program, the post-Fukushima lessons and the ASN requests has included the three important following Gen II PWRs upgrades in its PLE program:
— A strategy to allow corium stabilization without concrete basemat melt-through by the molten core after RPV failure;
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— A strategy to remove heat from the containment without containment venting;
— A strategy to reduce the iodine release in case of containment venting during a core melt accident.
There are also few remaining issues associated to uncertainties in the studies, which are relevant for both the second and the third objectives: effects of an ex-vessel steam explosion, effects of an uncontrolled injection of non-borated water during accident progression (inherent heterogeneous dilution), optimization of water management to limit residual risk for hydrogen and increase the in-vessel corium stabilization possibility (in-vessel water injection during in-(in-vessel core degradation, spray system activation, …).
4. MODIFICATIONS TO AVOID REACTOR CONCRETE BASEMAT MELT-THROUGH BY THE MOLTEN CORE AFTER RPV FAILURE
To limit the risk of reactor basemat melt-through by molten core after RPV failure, EDF has retained the following strategies:
— The vessel cavity (or reactor pit) is modified to avoid any water entry before vessel failure (in the existing design, the spray system activation fills the cavity with water);
— The reactor sumps are filled with water before the vessel failure;
— In case of vessel failure after core melt, the corium falls and spreads in the dry vessel cavity and optionally in an adjacent area. After complete spreading, some triggers are passively activated, allowing water from the sumps to submerge the spread corium;
— This water allows the corium cooling and its stabilization.
After the review of the principle of these modifications, IRSN has highlighted that:
— This strategy reduce significantly the possibility of containment failure by steam explosion in a flooded vessel cavity and is a good compromise between efficiency and feasibility for the ex-vessel corium stabilization;
— The size of the corium spreading area, depending on the reactor, has to be further discussed;
— Some design features have still to be defined, such as the passive trigger system for water submerging above the corium into the reactor cavity;
— There is a need for instrumentation (for instance to know the sumps water level)to provide information to the emergency teams, in order to alert any situation where the spread corium would not be submerged by water (emergency teams shall anticipate any need for additional water into the sumps).
The final studies of these modifications will be analysed during the next steps of the safety review process.
5. MODIFICATIONS TO REMOVE THE DECAY HEAT FROM THE CONTAINMENT WITHOUT OPENING THE EMERGENCY CONTAINMENT FILTERED VENTING SYSTEM
In order to allow the possibility to remove the decay heat from the containment without opening the EFCVS, EDF intends to implement a disposal (Fig. 1) composed of:
— A fixed circuit (located in the fuel building for the 900 MWe series).
A pump qualified to extreme external hazards conditions and SA situations;
An injection line connected to the cold leg of the primary coolant circuit and another one feeding the sumps of the reactor containment building;
A suction line connected to the safety injection tank (direct injection) and another one pumping in the sumps of the reactor containment building (recirculation);
A heat exchanger;
Actuators enabling the disposal activation from the control room.
— A “cooling mobile circuit” (ultimate heat sink) composed of a mobile pump and hoses directly drawing up in the heat sink and lined on the heat exchanger by the EDF rescue team - FARN (Nuclear Rapid Response Force).
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After the review of the principle of this modification, IRSN has concluded that:
— The new disposal intended by EDF is satisfactory in principle and should enable to remove the decay heat from the containment, if the reactor cooling by the steam generators had previously been realized during a sufficient time;
— An appropriate instrumentation is important to avoid any excessive pressure or temperature for the new disposal circuit and for the containment building;
— The possibility of circuit leakage during operation in severe accident conditions has to be considered with specific provisions like instrumentation to detect leakage, contaminated liquid management, reliable isolation valves;
— FARN activation criteria are importantto install the mobile ultimate heat sink in due time with margins.
The final studies of this modification and its qualification to severe accident conditions will be analysed during the next steps of the safety review process.
6. MODIFICATION TO REDUCE THE IODINE RELEASE IN CASE OF FILTERED CONTAINMENT VENTING DURING A CORE MELT ACCIDENT
On the 900 MWe PWR series, in case of core melt accident, silver released from the control rods and deposited into the sumps water would enable iodine stabilization in the sumps. The 1300 and 1450 MWe PWRs control rods include a limited quantity of silver compared to 900 MWe so this stabilization process would be limited.
In order to reduce the gaseous iodine release in case of filtered containment venting during a core melt accident on the 1300 and 1450 MWe PWRs, EDF is installing some sodium tetraborate (borax) baskets on the sumps floor to passively alkalize the sumps water and consequently trap iodine.
After the review of this modification, IRSN has concluded that:
— Implementation of borax baskets will allow trapping a large part of the iodine in the reactor building sumps for accident where sodium hydroxide cannot be injected by the containment spray system;
— The release coming from the iodine in the upper part of the containment would not be fully affected by the iodine stabilisation in the reactor building sumps;
— There is still an interest to examine, for all Gen II PWRs, the possibility to upgrade the existing EFCVS to reduce the gaseous iodine release.
This topic and also the EFCVS seismic reinforcement for all Gen II PWRs will be discussed during the next steps of the safety review process.
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7. CONCLUSION
For IRSN, the principles of the modifications planned by EDF in the framework of PLE (and the Post-Fukushima action plan) will really give more credit in reaching a stable reactor state after a core melt without any major failure of the reactor containment. During the coming period, the EDF detailed studies supporting the demonstration of the efficiency of main Gen II PWRs upgrades will be carefully examined, using available knowledge from research programs, especially for the corium coolability during molten core concrete interaction.
This will allow the present French reactors to reach as far as possible the EPR safety objectives for severe accident, even if some specific EPR features cannot be directly implemented for the existing reactors (core-catcher with corium bottom and top cooling, buildings arrangement to prevent possibility of direct leaks to outside, …).
After reaching a demonstration of the efficiency of these new disposals, will come the industrial challenges associated to the practical implementation on each reactor.
REFERENCES
[1] RAIMOND, E., et al., “Progress in the implementation of severe accident measures on the operated French PWRs – some IRSN views and activities”, OECD/NEA Workshop on the Implementation of Severe Accident Management Measures (ISAMM 2009), Böttstein, Switzerland.
[2] Technical Guidelines for the Design and Construction of the Next Generation of Nuclear Power Plants with Pressurized Water Reactors, adopted during the GPR/German experts’ plenary meetings held on October 19th and 26th 2000.
[3] CÉNÉRINO, G., et al., “Radiological objectives and severe accident mitigation strategy for the generation II PWRs in France in the framework of PLE”, Third International Conference on Nuclear Power Plant Life Management (PLiM) (Proc. Int. Conf. Salt Lake City, Utah, USA, 2012).