Loviisa Nuclear Power Plant, two 510 MWe (gross) VVER-440 units, is owned and operated by Fortum Power and Heat Oy (former Imatran Voima Oy). The company name was changed in 1999 after the merging of Imatran Voima Oy and the petrochemical company Neste Oy into Fortum Corporation.
In 1999 the gross production of Loviisa 1 was 4066 GWh and the capacity factor was 91.0% (gross). The annual refuelling and maintenance outage lasted 19 days. The gross production of Loviisa 2 was 4165 GWh, the capacity factor 93.2% and the length of the refuelling outage was also 19 days. The annual collective radiation doses were 0.80 manSv and 0.56 manSv for Loviisa 1 and Loviisa 2 respectively.
In the year 2000 Loviisa 1 produced 3798 GWh (gross), the capacity factor was 84.8% and the refuelling and maintenance outage lasted 44 days. Loviisa 1 had an inspection outage, which is performed every fourth year. In 2000 the gross production of Loviisa 2 was 4075 GWh, the capacity factor was 91.0%, and the refuelling outage lasted 19 days. The collective radiation doses in 2000 were 1.73 manSv for Loviisa 1 and 0.54 manSv for Loviisa 2.
Eight events in 1999 and seven events in 2000 were classified on the International Nuclear Event Scale (INES). In both years there was one level 1 event and the classification of the other events was 0. The first level 1 event was revealed in a periodic test during the refuelling outage of Loviisa 2. It was noticed that the identification tags of two manually operated valves in the nitrogen blowing (cleaning) lines of the emergency cooling system sump strainers, had been interchanged. In these two-redundant sump systems, the cleaning operations of the strainers, if needed, would in this case have affected the wrong redundancy. The other level 1 event occurred at Loviisa 1 during the refuelling outage in 2000 when recurrent line-up errors (in filling the suction line of the draining pump in the fuel pond cooling system) caused reactor pool water leakages onto the floor of the steam generator compartment.
In the first four months of 2001 three events have been classified on the INES scale. These events have been below the INES scale (level 0).
The first stage of the final repository for medium and low level radioactive waste was licensed in 1999. The first stage includes underground tunnels for solid low active waste. In 2001 STUK approved the Preliminary Safety Analysis Report of the solidification plant and detailed planning of the plant is going on. The civil construction works will be started later in 2001.
The interim storage for spent fuel had to be extended when the returning of spent fuel to Russia was terminated in 1996. The extension part of the storage was completed in 2000.
In 1998 Fortum Power and Heat Oy launched the environmental impact assessment procedure (EIA) of the new nuclear power plant. The planned location is the site of Loviisa 1 and 2. The EIA report was finalised in 1999 and the co-ordination authority, Ministry of Trade and Industry, gave its statement on the procedure in 2000.
The project for the modernisation and power upgrading of Loviisa NPPs gave an excellent possibility to take advantage of the latest development in the nuclear power plant technology. The key aspects were to verify the plant safety, to improve production capacity and to give a good basis for the extension of the plant's lifetime to at least 45 years.
In the first phase, before starting the project, a feasibility study for uprating of the reactor thermal power was carried out. The main result was in short that no technical or licensing issues could be found which would prevent the raising of the reactor thermal output up to 1500 MW from the original level of 1375 MW.
The carefully prepared feasibility study gave a good picture of the necessary plant modifications as well as essential areas in the analysis work, which was of use in planning the critical works and the time schedule of the project. The feasibility study focused on the following tasks:
The main objectives for the project were based on the feasibility study:
(1) Plant safety level as a whole will be checked and, if needed, improvements will be made.
(2) Plant units will be licensed for 1500 MW reactor thermal output.
(3) Gross electric output of the plant units will be raised to about 500 MW.
(4) Assistance to the life time extension of the plant units.
(5) The long-term availability of the plant is not impaired.
(6) Increase in the expert knowledge of staff.
The feasibility study concerning the reactor power upgrading and improvements of the turbine efficiency was started in spring 1994. After good results from the study, the preparation of the project plan began in summer 1995. Critical works in the time schedule, such as the revision of the Final Safety Analysis Report and the preparation of certain plant modifications, were started immediately.
The first step of the trial run by 103% reactor power could be started in January 1997. Test runs continued step by step during the year, and the last transient test by final reactor power was completed successfully in December 1997.
The Council of State awarded a new operating license for Loviisa NPP in April 1998. The license is valid until the end of 2007 for 1500 MW reactor thermal power, which is 9.1% more than the previous power level of 1375 MW. Measures to improve the efficiency of the steam turbines will continue in the annual maintenance outages until the year 2002.
The implementation of the project was carried out in co-operation between Loviisa NPP and Fortum Engineering. In addition, many other organisations such as the Technical Research Centre of Finland (VTT) participated in the work. Special attention was paid to the QA routines in the project as well as to the co-ordination of the work in several organisations. One example of this was the particular subject-specific specialist groups which were established to overview essential sections such as nuclear safety and commissioning.
The work was divided into the following ten sub-projects each having a responsible person from the organisations of both Loviisa NPP and Fortum Engineering:
(1) Operating licenses
(2) Other licenses
(3) Safety analyses and basic data management
(4) FSAR revision and comparison of the plant with regulatory body guidelines
(5) PSA (including level 2 PSA)
(6) Modification of the turbines
(7) Electricity systems
(8) Reactor and fuel
(9) Process systems and automation
(10) Commissioning and revision of instructions.
Increasing the electrical output by about 50 MW at each unit was part of the Loviisa modernisation programme. After completing the upgrading of the reactor thermal output in April 1998, more than 80% of the total increase in the electrical output was fulfilled. The rest of the power increase is available when the measures to improve the steam turbines are completed in the year 2002.
The reactor power upgrading from 1375 MW to 1500 MW was planned on the basis of optimising the need for heavy plant modifications. In the primary side and the sea water cooling system, the mass flow rates were not affected, but the temperature difference has been increased in proportion to the power upgrading. In the turbine side, the live steam and the feedwater flow rate were increased by about 10%; the live steam pressure was not changed.
The reactor fuel loading was considered on the basis of the previous limits set for the maximum fuel linear power and fuel burn-up. The increase in the reactor thermal output was carried out by optimising the power distribution in the core and the power of any single fuel bundle was not increased above the maximum level before power upgrading. In parallel with this work, more advanced options related to the mixing rate of the cooling water in the fuel subchannels and the increasing of fuel enrichment were investigated. The dummy elements installed on the periphery of the core in Loviisa 1 and 2 were preserved to minimise irradiation embrittlement of the reactor pressure vessel.
The VVER 440 design margins in the primary side are rather large and the hardware modifications needed there were quite limited. Replacement of the pressuriser safety valves was indicated already during the feasibility study as a necessary measure because of the power upgrading. Most of the other substantial measures in the primary side were carried out on the basis of the continuing effort to maintain and raise the safety level of the plant, and they were not directly included in the power upgrading.
It was necessary to carry out more extensive measures in the turbine plant and to the electrical components. Steam turbines were modified to a higher steam flow rate. Because of these measures, also the efficiency and operation reliability has improved. Certain modifications were carried out in the electrical generators and the main transformers to ensure reliability in continuous operation with the upgraded power output.
The last step in the process to upgrade the reactor thermal power was the long-term trial run to verify the main process parameters as well as plant operation in both steady state and transient situations. The trial run was carried out at gradually upgraded reactor power with a power level of 103%, 105%, 107% and finally 109%. Transient tests defined in the test programme were performed with a reactor thermal power of 105% and 109%. The test results correspond very well with all analyses and calculations. All the acceptance criteria for the tests were fulfilled.
The modernisation programme as a whole was started from the basis of the positive safety progress. This was applied by taking advantage of the latest development in calculation codes and technology as well as feedback of the operating experience, expertise in the ageing processes and safety reassessment coupled with the evolution of safety standards.
STUK was closely involved at every stage of the project, from the early planning of the concept to the evaluation of the results from the test runs. STUK examined all the modification plans that might be expected to have an impact on plant safety. Individual permits were granted stage by stage, based on the successful implementation of previous work.
The renewal of the operating license for the increased reactor power was carried out in the following steps:
The environmental impact has been assessed in the EIA Report, which was completed in December 1996. This was the first time in Finland (parallel with TVO plant having a corresponding modernisation programme) the EIA Procedure has been applied to a nuclear power plant. The law and the decree set certain procedures, including a public hearing for screening, scoping and the EIA statement, which are the stages of this procedure.
The result was that the reactor thermal power upgrading has no other considerable environmental impact than a slight increase in the outlet temperature of the cooling water. This means that the maximum temperature increase of the cooling water in the main condenser, before released back to the sea, is about 1°C higher than the previous temperature increase, which was typically close to 10°C.
An extensive safety review and comparison of the plant with the latest national regulatory body guidelines (YVL guides) have been carried out. This work was performed taking into account many international standards, such as the IAEA standard "A Common Basis for Judging the Safety of Nuclear Power Plants Built to the Earlier Standards INSAG-8". As a result of the work, a particular safety review report has been completed.
A part of the safety review and the licensing process of the reactor power upgrading was the renewal of the Final Safety Analysis Report. New accident analyses have been made concerning the containment pressure, LOCA and MSLB, for example. In addition to the accident analyses, there is a large number of transient situations that have also been analysed. The risk for a radioactive release to the environment was probabilistically considered (PSA level 2) for the first time for Loviisa NPP.
The Loviisa severe accident program, which includes plant modifications and severe accident management procedures, was initiated in order to meet the requirements of the Finnish regulatory authority, STUK.
Fortum's approach for severe accident assessment and management for Loviisa is based on four successive levels. The first level of the approach is to ensure that severe accidents can be prevented with high probability. The quantitative targets for the overall core damage frequency (CDF) obtained from PSA level 1, are 10-4 /reactor year for existing plants.
The second level is to show a very low fraction of overall CDF for those classes of accident sequences which can be assumed to directly lead to a large release. Such sequences are the ones with an impaired containment system function, high pressure core melt sequences and reactivity accidents leading to core damage. The class called sequences with impaired containment function consists of containment by-pass sequences (especially, primary to secondary leakage accidents), sequences with pre-existing openings, containment isolation failures, containment pressure suppression system by-passes and sequences with induced leakage outside the containment.
On the third level of the approach, the focus is on physical phenomena capable of threatening the containment integrity. The challenge to the containment integrity due to any physical phenomena should be excluded either by excluding the phenomenon itself as physically unreasonable or by showing that the loads caused by the phenomenon are tolerable. The phenomena considered include in-vessel and ex-vessel steam explosions, hydrogen burns, direct containment heating, missiles, slow overpressurization due to steaming and generation of noncondensable gases, core-concrete interaction, recriticality of the degraded core and core debris, and temperature loadings of the containment. It is obvious that plant specific studies are needed for proper treatment of the individual phenomena. Instead of traditional PSA level 2 type of approach, in case of Loviisa, Fortum has treated the main phenomenological, Loviisa-specific questions along the lines of the ROAAM (Risk Oriented Accident Analysis Methodology) approach.
After successful exclusion of the containment system and structural failures, the fourth and final level of the approach is to define the radioactive releases through containment leakages. The releases during the managed accident sequences should stay below the acceptable criteria concerning acute health effects and land contamination.
For Loviisa, the approach translates to ensuring the following top level safety functions:
The cornerstone of the SAM strategy proposed for Loviisa is the coolability of corium inside the reactor pressure vessel (RPV) through external cooling of the vessel. Since the RPV is not penetrated, all the ex-vessel phenomena such as ex-vessel steam explosions, direct containment heating and core-concrete interactions can be excluded. The only energetic phenomena remaining which could have potential to threaten the containment integrity are hydrogen burns.
Some of the design features of the Loviisa Plant make it most amenable for using the concept in-vessel retention (IVR) of corium by external cooling of the RPV as the principle means of arresting the progress of a core melt accident. Such features include
On the other hand, if in-vessel retention was not attempted, showing resistance to energetic steam generation and coolability of corium in the reactor cavity could be laborious for Loviisa, because of the small, water filled cavity with small floor area and tight venting paths for the steam out of the cavity.
The main focus of the thermal studies for IVR is therefore on finding out 1) the actual heat flux from the molten corium pool and 2) the critical heat fluxes at the corresponding locations on the RPV wall. Because of the relatively thick RPV wall, and because of the crust, which creates isothermal boundary conditions for the molten pool, the in-vessel and ex-vessel heat transfer phenomena can be effectively decoupled from each other.
An extensive research program was carried out by Fortum. The work included both experimental and analytical studies on heat transfer in a molten pool with volumetric heat generation and on heat transfer and flow behaviour at the RPV outer surface.
Based on experiments, the IVR concept for Loviisa was finalised in April 1994. The concept includes plant modifications at four locations. The most laborious one is the modification of the lower neutron and thermal shield such that it can be lowered down in case of an accident to allow free passage of water in contact with the RPV bottom. Other two modifications include slight changes of thermal insulations and ventilation channels in order to ensure effective natural circulation of water in the channel surrounding the RPV. Finally a strainer facility will be constructed in the reactor cavity in order to screen out possible impurities from the coolant flow and thereby prevent clogging of the narrow flow paths around the RPV.
The conceptual design was submitted to STUK for approval and approval in principle was received in December 1995.
Based on plant-specific features, the only real concern regarding potential energetic phenomena is due to hydrogen combustion events. The Loviisa reactors are equipped with ice-conderser containments, which are relatively large in size (comparable to the volume of typical large dry containments) but have a low design pressure of 0.17 MPa. The ultimate failure pressure has been estimated to be well above 0.3 MPa. An intermediate deck divides the containment in the upper (UC) and lower compartments (LC). All the nuclear steam supply system (NSSS) components are located in the lower compartment and, therefore, any release of hydrogen will be directed into the lower compartment. In order to reach the upper compartment, which is significantly larger in volume, the hydrogen and steam have to pass through the ice-condensers.
Because of the relatively low design pressure of the containment, the hydrogen burns that can create a potential threat include not only detonations, but also all large-scale combustion events that are rapid enough to yield an essentially adiabatic behaviour. An additional concern, which is caused by the type of the containment, occurs when the steam and hydrogen mixture passes through the ice-condenser. The steam will be condensed in the ice beds, which could potentially lead to very high local hydrogen concentrations.
In the early 1990's an extensive research program was initiated at Fortum to assess the reliability and adequacy of the existing igniter system. One of the focus areas in the studies was to determine the prerequisites for creating and maintaining a global convective flow loop around the containment for ensuring well mixed conditions. The global flow loop which passes from the lower compartment through an ice-condenser to the upper compartment and back to the LC through the other ice-condenser is necessary in order to bring air into the LC and thus to be able to recombine or burn hydrogen in a controlled way already in the LC. The experiments and the related numerical calculations demonstrated that the global convective loop will be created and maintained reliably provided that the ice-condenser doors will stay open.
The studies have been completed and the new hydrogen management strategy for Loviisa has been formulated. The new hydrogen management scheme concentrates on two functions: ensuring air recirculation flow paths to establish a well-mixed atmosphere (opening of ice condenser doors) and effective recombination and/or controlled ignition of hydrogen. Plant modifications which are necessary include the new hydrogen recombination devices and a dedicated system for opening the ice-condenser doors.
The studies on prevention of long term overpressurization at Loviisa started by considering the concept of filtered venting, as was done for many European NPPs after the Chernobyl accident. However, the capability of the steel shell containment to resist subatmospheric pressures is poor. If using filtered venting, it is possible that the amount of noncondensable gases after the venting is significantly less than originally, which later--after cooldown of the containment atmosphere--may lead to subatmospheric pressures and possibly collapse of the containment. Therefore, alternative solutions were sought for.
Since the concrete used in the reactor cavity of Loviisa does not contain any CO2, the amount of noncondensable gases (except for hydrogen) generated during core-concrete interaction would be practically zero. Therefore, the overpressure protection of the containment could be limited to condensing the steam produced. An obvious way of doing this is to spray the exterior of the containment steel shell. Later on, the concept of in-vessel retention was introduced to Loviisa (as discussed above), which excludes core-concrete interactions altogether and thus finally ensures that no noncondensable gases apart from hydrogen need to be considered.
The system is designed to remove the heat from the containment in a severe accident when other means of decay heat removal from the containment are not operable. Due to the ice condenser containment, the time delay from the onset of the accident to the start of the external spray system is long (18-36 hours). Thus the required heat removal capacity is also low, only 3 MW (fraction of decay power is still absorbed by thick concrete walls). The system is started manually when the containment pressure reaches the design pressure 1.7 bar. Autonomous operation of the system independently from plant emergency diesels is ensured with dedicated local diesel generators. The single failure criterion is applied. The active parts of the system are independent from all other containment decay heat removal systems. There are no active parts of the system inside the containment.
The both units Loviisa 1 and 2 have their own external spraying circuits and spray water storage tanks. The cooling circuit of the spraying system and the dedicated diesel generators are common for both units. The ultimate heat sink is sea water.
The design calculations were carried out with Fortum's own simplified containment thermal-hydraulic code PREDEC. The PREDEC calculations were supported by experiments carried out at the HDR containment (tests E11.2 and E11.4) in Germany. These experiments were aimed at studying the hydrogen distribution during stratified conditions inside the containment. The main result from the HDR experiments was that the PREDEC code could be used for the design calculations of the external spray system.
The influence of the external spray system was further studied experimentally using the VICTORIA facility.
The primary depressurization is an interface action between the preventive and mitigative parts of SAM. If the primary feed function is operable, the depressurization may prevent the core melt. If not, it sets in motion the mitigative actions and measures to protect the containment integrity and mitigate large releases.
Manual depressurization capability has been designed and implemented through motor-operated relief valves. Depressurization capacity will be sufficient for bleed&feed operation with high-pressure pumps, and for reducing the primary pressure before the molten corium degrades the reactor vessel strength. Depressurization is to be initiated from indications of superheated temperatures at core exit thermocouples. The depressurization valves were installed at the same time with the replacement of the existing pressurizer safety valves in 1996.
The SAM-strategy described in the previous chapters has lead to a number of hardware changes at the plant as well as to new severe accident guidelines and procedures.
The containment external spray was implemented at the two units in 1990 and 1991. Primary system depressurization capability was installed at both units in 1996. The major backfittings related to external coolability of the reactor pressure vessel and to opening the ice-condenser doors are, for the most part, implemented at Loviisa 1 in 2000 and at Loviisa 2 in 2002. Test samples of the new hydrogen recombination devices have been aged and tested in plant conditions and the devices will be installed in 2002. In addition to the mechanical equipment, the implementation includes also a new, dedicated, limited scope instrumentation and control system for the SAM-systems, a dedicated AC-power system and a separate SAM control room which is common to both units.
The severe accidents guidance for the operating crew consists of SAM-procedures for the operators and of a so-called Severe Accident Handbook for the Technical Support Team. The SAM procedures are entered after a prolonged uncovery of the reactor core indicated by highly superheated core exit temperatures. The procedures are symptom oriented and their main objective is the protection of containment integrity through ensuring the top level severe accident safety functions. The most important operator actions after the core uncovery are the ensuring of containment isolation, primary circuit depressurization, opening of ice-condenser doors in order to ensure mixing of hydrogen, lowering of the neutron shield of the lower part of the RPV and, in the long term, starting of the containment external spray. The Severe Accident Handbook contains background material for the procedures and it should facilitate the support team in gaining understanding of the progress of the accident and of potential means of recovery.
See Annex III, Development of the safety of the Olkiluoto nuclear power plant and Chapter 4. The text in Annex III applies also for Loviisa nuclear power plant. The only difference is that the situation of Fortum's inspection qualifications (last paragraph) is that the qualification of in-service inspections based on ENIQ recommendations was started in 1998.
VVER reactor operating experience is collected, screened and evaluated by a dedicated operating experience feedback group composed of engineers from the plant operation organisation and from Fortum Nuclear Services. The group can give recommendations on further studies and measures to the operating organisation. The main information to be handled comes from WANO Moscow Centre which links all the VVER reactor operators. Additional reports are received from the IAEA, OECD/NEA and NRC, and naturally the activities of the operation experience feedback group are not limited only to VVER reactors.
The plant managers of VVER-440 reactors run a so-called VVER Club with periodic meetings. The plant operation problems, modernisation, back-fitting, lifetime extension and safety questions are handled and experiences are exchanged in these meetings and in further individual contacts.
Loviisa Power Plant participates in the WANO Peer Review Programme by sending peers to other plants including VVER plants. In February - March 2001 WANO Moscow Centre organised a Peer Review at Loviisa Power Plant. Several peers including the team leader came from other VVER plants. This co-operation between plants of the same design serve also the exchange of relevant operation experiences.
Fortum Nuclear Services has been a partner in several international and Finnish safety and quality related support programmes. Loviisa Power Plant has participated in some of these projects and has had a possibility to widen the organisation's experience on current development with other VVER operators. The same applies to a couple of direct commercial consultation projects which have been managed by Loviisa Power Plant.
In the area of radiation protection, Loviisa Power Plant is participating in the IAEA Technical Co-operation Project RER/9/063 "Enhancing Occupational Radiation Protection in Nuclear Power Plants". The workshops organised by the project bring together VVER and RBMK health physicists to exchange information on current issues in radiation protection.
An analysis of reported events often reveals that deficiencies of modification management have been a contributing factor. Such deficiencies include late planning, lack of co-ordination with other works, last moment changes, documentation defects, unfinished disassembling works and delayed updating of the documentation.
Proper planning and scheduling are the key factors in modification management. Loviisa Power Plant has completed an extensive project training course in 2000 for those in the operating organisation who will be involved in future modification projects. Successful projects such as the plant modernisation and power upgrading have been used as good examples.
The scheduling of the modification planning for the next outage is fixed in order to get enough time for preparations. Minor modifications are concentrated to every second annual maintenance outage and major works are carried out every fourth year. This is accomplished by starting from a long term investment planning which converts into a long term modification plan.
During the maintenance outage the scheduling office is directing their efforts from the previous control of the overall schedule to controlling the individual work packages including also the modification works. In the main schedule more time is allocated to tests related to start-up. New arrangements for handling the work orders in the main control room have been introduced for the next annual outages. The idea is to even up the work load in the main control room and decrease the disturbance of the operators.
Quality procedures for executing modifications have recently been updated. The authority to make decisions on last moment changes in the scope or schedule of the modification works has been clarified.
After Fortum Corporation was formed a need for an updated quality policy was obvious. In 1999 a quality statement "Fortum's Policy Commitment to Quality in the Nuclear Power Operations" was issued by the president of Fortum Power and Heat Oy. The statement has been confirmed in 2001 also by the new management of Fortum Power and Heat Oy.
The recent development of the plant quality management system is based on the principle of continuous improvement in accordance with the observations and remarks made in quality audits and quality assessments.
Loviisa Power Plant adopted in 2001 a newly formulated management procedure which defines an annual planning process from strategic planning to annual reports. A first 10-year strategic plan for the power plant was developed in 2000.
A second important and new procedure describes those review processes (e.g. management reviews, self assessments), which are needed in an effective quality management system.
In the internal quality audits, new efforts are directed to the evaluation of the recurrence of events. These have considerably increased the necessary background work both in the preparation and in the reporting phase of an internal audit.
An evaluation of the plant quality management system against the ISO/DIS 9001, 9004:2000 standards was made in 2000 by Fortum Engineering. The work continues in 2001 and a similar comparison with IAEA Safety Series No. 50-C/SG-Q has already been ordered.
Preparation of the environmental management system according to the ISO 14001 standard is included in the quality management system. Preparations at the procedure level have introduced a new chapter in the Quality Manual and in the updating of numerous quality procedures. A novel environmental aspect shall be considered in internal audits and new part-time auditors have been trained for environmental evaluations. The readiness for certification of the environmental management system should be achieved by the end of 2001.
The present tracking system for quality and safety decisions, obligations and actions has capacity limitations and a new tailored application will be delivered in 2001.