Answers 1

Answers to the questions and comments related to the Finnish report “Compliance with the obligations of the Convention on Nuclear Safety”

Finland has received questions and comments from the following countries: Austria, France, Hungary, Japan, Mexico, Russia, Slovenia, Sweden, Ukraine

Article 6 – Existing Nuclear Installations

Year 2000 Computer Issue
How does Finland ensure that each licensee has an adequate strategy and action plan in place to deal with the year 2000 safety issue?

Discussions on the projects for dealing the year 2000 issues started between STUK and both of the Finnish utilities in 1997. At that time both utilities had already Y2K projects underway. At the first quarter of the year 1998 STUK sent a letter to the utilities asking for the project plans and the status reports. In the status report it was asked to report the computer based systems and equipment regardless their safety classification.

Both of the Finnish utilities have delivered to STUK their project plans including the major tasks, responsibilities and milestones. The projects cover computer systems of the power plant, systems used in laboratories, electric and automation systems near the process utilising firmware or programmable logic, communication and data transfer systems, construction automation and maintenance systems. The utilities are required to report the progress of the projects regularly. Based on the project plans and further status reports, inspections related to the project documentation and management have been made at the NPP sites during the year 1998. STUK has also attended the testing of some systems such as plant process computer.

The Loviisa plant completed on-site and laboratory testing of all safety relevant systems already before the end of 1998. Some equipment and software changes remain to be done in 1999, as well as testing of a few non-nuclear safety systems. At the Olkiluoto plant more work is left for 1999.

During 1999 there will be two audits made to the both of the NPP sites by STUK. The first one March–April, and the second during October. The utilities are also required to deliver a safety assessment to justify the continuous operation of the power plant during the roll over of the Y2K-critical dates and a contingency plan, which is applied during critical roll over periods. The contingency plans include evaluation of all external connections and supplies to the NPP's, and measures to be taken account should there be cuts in any of them.

The Report establishes that Section 11, 12, 17 and 18 are not complete, but Section 28 – Decision 395/1991 and plant modifications to be implemented in the future, considered that this Decision has already been satisfied. How is his Decision considered satisfied if actions still in process?

The Report refers to the conclusions of the safety review of STUK concerning the Loviisa plant. The safety review was carried out in connection with the re-licensing of the Loviisa plant.

Section 28 of Decision 395/1991 provides some exemptions from the provisions of Sections 11, 12, 17 and 18 of Decision. These exemptions can be accepted for nuclear power plants that were in operation before the Decision was issued. Section 28 also refers to Section 27, which requires such actions for further safety enhancements which can be regarded as justified considering operating experience and the results of safety research as well as the advancement of science and technology. The technical solutions of the plant in question shall also be taken into account. Accordingly, there are planned modifications to be implemented in the near future at the Loviisa plant e.g. for severe accidents. So the situation is in compliance with the Decision 395/1991. This issue is discussed in more detail in Annex 5 of the Report.

Article 7 – Legislative and Regulatory Framework

The Finnish legislative and regulatory framework seems to be well structured and very comprehensive. Detailed safety requirements exist in the YVL Guides as rules to be complied with unless some alternative acceptable procedure or solution can be presented by the licensee. How prescriptive is this system in practice? How often do the licensees present alternative solutions to safety problems? What flexibility does the system allow the licensees in their development of safety?

In Finland we do have a quite comprehensive system of regulations and safety guides. This system is surely more prescriptive than in many other countries. However, we don't consider the Finnish regulatory system to be too prescriptive in the sense that it would somehow restrict or limit neither the development of safety nor the safety culture of the licensees.

Much of the contents of the YVL Guides is guidance on how to present various issues to STUK for review (format, contents and schedules for documents to be submitted). Most of the technical requirements are general goals and acceptance criteria rather than descriptions on how to meet the goals. The YVL Guides also remind on issues that must be considered, but do not say how to treat those issues. At the most prescriptive end, there are requirements such as the minimum level of redundancy and diversity of specific safety functions.

The procedure to establish YVL guides is such that the licensees are always given the opportunity to comment the draft guides. Also meetings with the licensees are arranged to discuss new safety requirements and their basis. The involvement of licensees in preparing the safety guides is therefore strong; however, of course no acceptance of new safety requirements by the licensees is required. All YVL Guides also undergo a separate review by Advisory Committee on Nuclear Safety.

The regulatory strategy in preparing the YVL Guides is such, that they give the best means known to STUK to ensure the safety. The YVL guides give the minimum safety requirements and not the maximum ones. Therefore, they don't limit the licensees' actions to enhance the safety. Also alternative solutions providing the same safety level are acceptable.

In practise, because of the involvement of the licensees in the preparation of the guides, the number of presentations of alternative solutions to safety problems have been very low. Examples of such alternative solutions are found in the safety classification of equipment, and in the training and qualifications of the operating organisation.

See also the answer to the next question.

Who authorised the STUK to issue YVL guides? To what extent should these guides be regarded obligatory to the operator?

In 1970's, when the first YVL Guides were issued, the authority was loosely based on conditions in the Construction Licenses (a general authority to give orders as needed to ensure safety). Now the authority is firmly and explicitly based on the Nuclear Energy Act, Section 55. The authority to issue detailed requirements is also mentioned in the Decisions 395/1991, 396/1991, 397/1991 and 398/1991 concerning safety, physical protection and emergency preparedness of nuclear facilities.

YVL Guides are rules an individual licensee or any other organisation concerned shall comply with, unless some other procedure or solution is presented to STUK by which the safety level provided in an YVL Guide is achieved. Based on Sections 27 and 28 of Decision 395/1991, STUK may decide that some new requirements are not applied to a nuclear power unit in operation. In this consideration, operating experiences, safety research results and the advancement of science and technology as well as the technical solutions of the plant are taken into account. In connection with the re-licensing of the Loviisa and Olkiluoto plants, also a comprehensive evaluation was carried out on how the requirements of YVL Guides are complied with. This evaluation resulted in many corrective actions as well in exemptions from specific requirements.

It is not clear whether the parliament rejects or not the licensee's applications. Please explain if the licensee has alternatives to modify the application or whether it is decisively rejected.

The Parliament is involved only in the very first phase of a project for a new nuclear facility. The Parliament can only approve or reject a Decision in Principle issued by the Council of State (Government). Decision of Principle is required before a Construction License can be applied for. No modification of the application is possible when the decision of the Parliament has been done. However, the entire process can be started with a new application.

Article 8 – Regulatory Body

Which reporting obligations exist for STUK to different governmental organisations and the public?

STUK has e.g. the following obligations to report on its activities:

  • Sections 61-65 of the Decree on the state budget require an annual report on the activities, including the financial statements; in these reports it is described how the performance objectives and goals written in the annual "result agreement" with the Ministry of Social Affairs and Health were achieved. The report is submitted to the Ministry of Social Affairs and Health and to many other ministries as well as to the State Treasury.
  • Section 121 of the Nuclear Energy Decree requires an annual report on the regulatory control of the use of nuclear energy; the report is submitted to the Ministry of Trade and Industry.
  • Section 1 of the Decree on STUK requires that STUK informs on radiation and nuclear safety issues.

Based on its own initiative, STUK prepares e.g. quarterly and annually reports on

  • the use of nuclear power plants
  • emergency preparedness, and
  • environmental radiation monitoring.

STUK also prepares annual reports on the use of radiation. In addition, STUK publishes annually a general report describing the work of STUK. All these reports are available to various governmental organizations and to the public.

All the significant incidents in the area of radiation and nuclear safety are reported promptly.

Please provide more details about how the STUK inspection programme is planned and structured? In what level of detail does STUK inspect the nuclear installations? How much do inspections address safety culture and QA as compared to technical issues? Are inspection reports public documents?

Regulatory inspections relating to the operation of nuclear power plants can be put in two categories:

  • the periodical inspection programme (periodical inspections), and
  • the inspections which the power company is obliged to request in connection with measures carried out at the plant, or which STUK conducts at its discretion (specific inspections).

Examples of the specific inspections are inspections of repair and modification plans of vital safety equipment (including inspection of the QC results and actual equipment after work has been completed), assessment of the annual in-service inspection programme and its results, review of reactor refuelling plan and related analysis of core and fuel behaviour, examinations to verify the competence of key personnel, and verification of plant readiness to start-up after refuelling outage.

Periodical inspection programme has been reviewed and revised in 1998. The review was based on the experience gained from about 15 years implementation of the previous programme. In 1999, inspections will be carried out according to the new programme.

The new inspection programme consists of three categories:

A. Management
A1 Safety management (once in two years)
B. Main work processes (including QA, training,...) (annually)
B1 Assurance and development of safety
B2 Operation
B3 Maintenance
C. Topical inspections (annually)
C1 Performance of safety functions and safety systems
C2 Electric and I&C
C3 Mechanical engineering
C4 Civil construction and structures
C5 PSA and utilisation of failure statistics
C6 Data administration
C7 Chemistry
C8 Management of nuclear waste
C9 Radiation protection
C10 Fire protection
C11 Emergency preparedness
C12 Physical protection

The Conclusions of each inspection (periodical or specific) shall be based on evidence acquired during the inspection. Level of details to be inspected depends on the characters of the inspection. Generally, periodical inspection programme is focusing on the work processes and practices (some results are sampled for getting evidence) where as the main purpose of specific inspections is to verify fulfilment of requirements. In general, STUK inspects structures, systems and devices, which are important to safety.

The aim of the revised inspection programme is to get an overall picture of the existent safety culture. This will be done annually by reviewing in integrated form the results (as documented in inspection reports) of the implementation of the inspection programme. Assessment of the effectiveness of the licensee's QA programme is included in all inspections, especially inspections A1 and B1-B3. A rough estimation between “software” and “hardware” issues is 50%/50%.

Inspection reports related to the new periodical inspection programme are public documents.

Which are the strategic objectives of the research and development programme of STUK and by what means are they realised in the current plans?

Strategic objectives of the nuclear safety research and development programme of STUK are based on safety regulations in the Decision of the Council of State (395/1991). In practise this means that STUK prepares every year a new safety research plan, taking into account operational experiences of NPPs, results of other safety research activities and also the development of technology. The research plan is divided into following categories:

  • Research and work related to the development of STUK inspection activities
  • Research and comparison analysis supporting the decision making of STUK
  • Research related to the development of expertise and also inspection and safety analysis preparedness in the following areas
    • Structural safety
    • Reactor safety
    • Safety management.
    • Waste management

The annual safety research plan is financed by STUK.

In addition to this annual activity, STUK participates in preparation of the nationally co-ordinated safety research programmes e.g. on operational safety, structural safety and waste management. These national research programmes last typically four years and are publicly funded.

The newest plan in the area of nuclear safety is called “The Finnish Research Programme on Nuclear Power Plant Safety (1999–2002) FINNUS”. The Technical Research Centre of Finland (VTT) coordinates the programme and also performs most of the research. Experts from STUK have participated in preparing detailed project descriptions and goals and STUK has nominated its own experts and members in the reference groups. The steering group of FINNUS-programme is chaired by STUK.

As mentioned in the CNS-report from Finland, Annexes 5 and 6 include more information on the safety research performed by the licensees.

According to Figure 1 and the text on page 11 the connection of STUK to the State Council is not unambiguous. What are the roles of Ministry of Trade and Industry, and those of the Ministry of Social Affairs and Health in that context?

In Finland all governmental institutions, such as research centres and regulatory authorities, are operating under some ministry. STUK is administratively under the Ministry of Social Affairs and Health. This means that it gets its annual budget from the funds allocated to the disposal of that Ministry, and it makes annually an agreement on expected work results with the Ministry. Furthermore, the Ministry prepares the appointment of STUK's Director General (final decision and appointment is made by the President of the Republic). In the substance matters STUK is independent, and can make all decisions relevant for nuclear regulation without any communication with the Ministry of Social Affairs and Health.

As concerns the Ministry for Trade and Industry, STUK has no administrative dependence on it. The Ministry for Trade and Industry is the administrative body with a task to prepare nuclear licensing and legislation matters for the Council of State (Government) which makes the final decisions. Such decisions are made e.g. on start-up of a new project for constructing a nuclear facility, on Construction and Operation Licenses, and on general safety regulations which are mandatory. However, the Ministry for Trade and Industry cannot put forward a license decision, if STUK has not given a positive statement on the safety aspects of the planned decision.

From the report, it is very clear that STUK receives technological support from VTT. Does VTT provide some kind of support to licensees? If so, please provide additional information of how the independence from one decision affecting both organisations is assured.

The problem of independence and conflicts of interest, when VTT carries out R&D for both the utilities and STUK, has been deeply discussed on several occasions. For example, in the beginning of 1998 an international evaluation was carried out for the national reactor safety research programmes RETU and RATU2 and this issue was taken up and discussed by the evaluators. They concluded that all the parties involved were very well aware of the potential problem and had taken different actions for handling the situation. Their overall conclusion was that there was no evidence that this is a cause for concern.

In practice following practical methods have regularly been used to create independence and avoid conflicts of interest:

  • This problem is recognised in VTT´s quality assurance and operating procedures and rules and in the framework agreement between STUK and VTT. Every person involved for contract work for utilities and STUK knows about the potential problem. VTT is also experienced in handling problems of similar type with competing industrial customers.
  • STUK, and the utilities as well, have also in-house expertise in all critical areas. Hence, the contracts for VTT often define the problem and methods for carrying out the analyses in a detailed way—and often the analysis work is carried out in close continuous interaction with STUK´s experts. The in-house expertise of STUK also provides independence in drawing conclusions from the results presented by VTT.
  • It happens sometimes that VTT cannot take a particular contract from STUK because of earlier related work for utilities. In these cases, STUK has used foreign contractors as necessary.
  • VTT is organised as a group of independent research institutes. In some cases it has been possible to agree that one VTT unit serves the regulator and another the industry.
  • If it is known well in advance that this type of problem is expected for a highly important issue, certain experts have explicitly been reserved for the use of one party only.

The independence aspect was also particularly addressed when follow-on national strategy for nuclear energy R&D was broadly discussed in 1998. The successor to the RETU and RATU research programmes, a combined programme called FINNUS, was designed so that the role of the STUK in the management and priority setting is even stronger than before. Utilities still directly participate in and sponsor the programme in areas where information of generic nature is produced, such as many international collaborative experimental projects, but plant specific applications of special interest for the utilities were separated into independent projects. On the other hand, most activities addressing future ALWR reactor concepts and some activities with immediate coupling with plant operations were regrouped into research programmes with utility driven management.

Article 9 – Responsibility of the Licence Holder

Has a Finnish licensee a right of recourse against his employees (in particular those in operational control), if they cause a nuclear damage either by their negligent behaviour or with the intent to cause such damage? If this is the case, is the right of recourse granted on the basis of the labour contracts entered into between the license holder and its employees or otherwise?

The Finnish licensee has a right of recourse against his employees in case of intentional damage. In case of other damage, a right of recourse would require that a "natural person" (employee) has in a written contract, which he has entered into with the licensee, specifically pledged to take responsibility for such damage.

Special provisions on the matter can be found in Section 21 of the Nuclear Liability Act (484/1972).

What are the amounts the license holder or its employees can be fined in accordance with Section 66 of the Nuclear Energy Act?

The monetary amounts of fines have not been set forth by legislation. General grounds for the assessment of a penalty under the regulations for the conditional imposition of a fine (1113/1990) are as follows:

  • quality and scope of the primary obligation (for the fulfilment of which the penalty is imposed)
  • solvency of the person liable to compensate
  • other relevant matters.

A practical basis for the assessment would be to impose a penalty, which is high enough to effectively promote the implementation of the primary obligation. In practice, penalties have never been imposed in Finland by virtue of the Nuclear Energy Act, and the question has never been seriously considered. The penalties are not regarded as a proper tool for promoting nuclear safety.

How heavy are the sanctions that can be imposed against the license holder or his employees for the criminal offences described in Section 69 of the Nuclear Energy Act?

The crime categories and scale of punishments in chapter 34 of the Penal Code are as follows:

  • endangerment offence injurious to health, scale from 4 months to 4 years of imprisonment
  • gross endangerment offence injurious to health, from 2 years to 10 years of imprisonment
  • criminal offence involving the use of nuclear explosives, from 2 to 10 years of imprisonment
  • negligent offence involving danger to the public, fine - one year of imprisonment
  • gross negligent offence involving danger to the public, from 4 months to 4 years of imprisonment.

Article 10 – Priority to Safety

The report states, "Section 4 of Decision 395/1991 provides that, an advanced safety culture shall be maintained when designing, constructing and operating a nuclear power plant."(p.13, paragraph 4). We understand that Finland is very conscious of safety culture. Please explain activities by the operating organisation to foster safety culture among its personnel.

Loviisa plant

The basic paper to study and to learn of the concept of safety culture are the IAEA publications: 75-INSAG-4 and IAEA-TECDOC, Ascot Guidelines. The development of the concept of safety culture within the nuclear and other industry areas is continuously followed. Other organisational behaviour knowledge is connected to safety culture concept. So such organisation culture is created and maintained, which gives priority to nuclear safety.

The basic safety culture seminar was held by STUK in 1991. At the same year, STUK had also assessed of safety culture at the Loviisa plant. The ASCOT seminar by IAEA/STUK was held in Finland 1993.

The concept of safety culture was introduced to the power plant management system in 1991 and the safety culture self-assessment of was performed in 1994-1995. The management group of the plant and the Plant Safety Committee reviewed the report carefully. On the basis of this assessment some improvements were made, e.g. the decision to include new training topics and targets to the training programs of separate groups, to the development plans and to the yearly QA-evaluation reports.

The plant quality policy emphasises safety culture and the matters manifesting safety. The safety culture has been part of basic and refreshing training since 1996. The separate training seminars, programs and courses support the learning and adopting the safety culture ideas. Safety culture is also emphasised in the articles in the plant bulletin.

The QA and Safety meetings at the plant chaired by the plant manager deal also with safety culture issues. The meeting is held about once a month. Management and QA staff review the quality system effectiveness annually. The safety culture matters are included in these reviews. The annual development plans include also some safety culture items. The safety culture is the permanent subject in the QA-report to the top management of the company twice a year.

Certain activities are important signs of good safety culture such as utilising operating experience, the development of training programs, QA-actions, cleanliness of the plant, the importance to follow procedures, the working atmosphere, the participation of the personnel. At the Loviisa plant these activities are often under discussion in separate connections and are developed continuously. The individuals are encouraged to report problems and mistakes.

Olkiluoto plant

Olkiluoto plant has emphasised the safety issues from the very beginning of plant operation. A certain part of this way of thinking was inherited from the vendor of the Olkiluoto plant, ABB Atom. After the Chernobyl Accident, the term “safety culture” was introduced and it has since been fostered in the following steps:

  • Familiarisation with the Summary report on the Post-Accident Review Meeting on the Chernobyl Accident, 75-INSAG-1, September 1986
  • Familiarisation with the report on Basic Safety Principles for Nuclear Power Plants, 75-INSAG-3, March 1988
  • Familiarisation with the report on Safety Culture, 75-INSAG-4, Feb 1991
  • Manager Seminar on Safety Culture, February 1991
  • Safety Culture course, February 1992
  • STUK's assessment of TVO's Safety Culture, January 1992
  • TVO's own assessment of its Safety Culture September 1992
  • Manager Seminar, September 1992
  • Chief Seminar on Safety Culture, January 1993 and February 1993
  • ASCOT-seminar, April 1993
  • Safety Culture presentation for TVO's Board of Directors, August 1993
  • Publication of “TVO's Safety and Quality Policy”-booklet, May 1995
  • Review of safety culture studies and assessment of TVO's activities against findings/recommendations made in Sweden (October 1995)
    • Hörmander study (Sydkraft)
    • Brodin study (Vattenfall).
  • Discussion of the above review results in TVO's
    • Safety Group
    • Management Group
    • Department meetings
  • Safety Culture lecture to managers and two lectures on Safety Culture to all staff of TVO given by the Director of nuclear safety of STUK, November 1998
  • Safety culture has been for several years a constant topic in the indoctrination training of all new employees, in the training given to contractor personnel working during outages and in the quality assurance training.

In what extent do the Finnish NPPs use safety indicators, incl. safety culture indicators, to monitor plant performance? Which indicators are used?

Loviisa plant

WANO (World Association of Nuclear Operators) Performance Indicators are used at the Loviisa plant. Exception is Thermal Performance, which has been dropped out in 1996.

These Performance Indicators are in use:

  • Unit Capability Factor
  • Unplanned Capability Lose Factor
  • Unplanned Automatic Reactor Scrams per 7000 Hrs
  • Safety System Performance
  • High Pressure Safety Injection
  • Auxiliary Feed Water
  • Emergency Diesel Generators
  • Fuel Reliability
  • Chemistry Index
  • Collective Radiation Exposure
  • Volume of Low-Level Solid Radioactive Waste
  • Industrial Safety Accident Rate.

Data has been gathered and reported to WANO since 1990 for all the above mentioned indicators (except Thermal Performance between 1990–1996, and today it is not used at all). Performance Indicator Reports since 1990 have been used to monitor the Loviisa plant performance & progress and to get additional perspective on performance relative to that of other power plants.

Other indicators are used at the Loviisa plant organisation, for example the Maintenance Group has an annual reporting system where the results of each supervisor in the whole group is monitored.

Also important system is a salary bonus system, which according to different indicators can mean an annual bonus in a magnitude of half a months salary.

Also there is a project to develop a quality indicator-system in the near future. It will consist of several different indicators from technical field to communication.

Olkiluoto plant

WANO Performance Indicators, including safety indicators, are used and reported to WANO. All of these indicators are reviewed also internally at least annually. Every indicator has one or several “owners” inside the utility. The “owner” is responsible for collecting and reporting to the internal co-ordinator the data required for the calculation of the indicator. The co-ordinator makes the validity check of the data, and reports it to the WANO Paris Centre. In addition, he prepares graphical presentation of the trends of the utility's indicators, presents it in the internal communication network to the staff and notifies the management of the company in case the trends cause concern. Annually he makes comparisons with the average values of the indicators from other nuclear power plants. The safety system component availability indicators are submitted to STUK, too.

In connection with the ASSET Self Assessment of the plant performance, October 1997 - September 1998, and ASSET Peer Review Mission in November 1998, safety culture was also evaluated according to the ASSET methodology. TVO has made a decision to continue the use of ASSET methodology as one tool in the event analysis work.

Which are STUK's requirements on independent safety review, of plant modifications etc., and which resources, in terms of staffing and competence, exist for independent safety review at the NPPs?

STUK's requirements regarding plant modifications are given in Guide YVL 1.8. According to these requirements, an independent safety review is needed for any licensee submittal to STUK related e.g. to plant modifications and reloadings. A normal practise is to have the licensee experts representing all relevant technical disciplines to review and concur the submitted documentation. A statement on the approval by the licensee is required also for all documents enclosed with the submittal, even if they originate from a third party (vendor, outside consultant, or other contractor). These requirements have been codified also in the licensee Quality Assurance System, guidance for which is given in Guide YVL 1.9.

Typically, this independent safety review at both nuclear power plants is managed by the personnel responsible for safety. In addition, the licensees have established separate PSA groups which may also participate in an independent safety review. Finally, higher-level oversight on safety is provided by the safety committees of the licensees, which are independent of the day-to-day work. However, these committees focus on more important or broad-scope matters and do not get involved with each modification performed.

The licensees also make use of external expertise to perform safety reviews, typically in the context of larger plant modifications. The ultimate responsibility for safety always rests with the licensee, however.

Inside STUK, the regulatory control of plant modifications always involves independent safety review. Third-party opinion or external review may be requested at STUK's discretion, but this is not required by any rule. It is STUK's policy to try to develop and maintain the needed in-house technical competence in order to be able to work independently and in a responsive fashion.

Article 11 – Financial and Human Resources

The Report does not contain concrete data on the number of plant staff and the staff qualification level. Please provide a more detailed overview of the workforce situation.

The organigram of TVO is attached (Annex 1). The total number of permanent plant staff is approximately 480 persons (about 60 with a university degree in engineering or natural sciences). In addition, there are about 100 hired people (for cleaning, guarding etc.). During the outages the number of external workforce is around 800–1200 people. The most important technical support organisation is the original plant supplier, ABB Atom Sweden.

The organigram of Fortum is attached (Annexes 2–4). The names of Neste and Imatran Voima (IVO) have been changed as a part of the development of energy group Fortum's operations. The new name of Fortum's subsidiary Neste is Fortum Oil and Gas Oy, a limited company corresponding to Fortum's Oil and Gas division. The new name of Fortum's subsidiary Imatran Voima is now Fortum Power and Heat Oy, a limited company corresponding to Fortum's Power and Heat division. The new names conclude the combination of IVO and Neste, started at the beginning of February 1998 with the establishment of IVO-Neste Group Ltd. The name of IVO-Neste Group was changed to Fortum in June 1998.

The Loviisa plant organigram is also attached (Annex 5). The total number of permanent plant staff is approximately 460 persons (about 30 with a university degree in engineering or natural sciences). In addition, there are 100 hired people (cleaning, guarding etc.). During the outages the number of external workforce is around 800 – 1000 people. The most important technical support organisation to the plant is the Fortum Engineering Ltd (earlier IVO International Ltd), where the number of persons in the nuclear area is around 120.

STUK has issued Guide YVL 1.7 to address the qualification requirements of plant staff taking care of safety related duties. The posts having special qualification requirements are the following: Plant responsible manager and his substitute, responsible managers for safeguards, physical protection and emergency preparedness as well as the plant operators. Also some testing and inspection personnel have to be qualified by the authorities.

The qualifications of other workforce are not especially controlled by the authority. However, it can be stated that the Finnish power companies do have written requirements for recruiting new employees. STUK reacts, if necessary, only if the performance of activities indicates that there is a problem with the level of competence of the staff.

Certain personnel categories at the NPPs are subject to STUK's approval. The approval is valid for 3 years. Which verifications are made by STUK as a basis for this approval, in particular the approval of the plants management group?

Approvals for 3 years are given only to shift supervisors and operators, who are required to keep their skills up-to-date. However, the first shift supervisor's or operator's license is valid for two years at most. Other approvals have no expire date. STUK participates oral examination of each operator. STUK also participates with its own questions in written examinations for the first license.

Approvals needed

The person appointed as a responsible manager or as his deputy must have an approval from STUK for this job.

The operator of the facility systems in the main control room must have an approval granted STUK for the job.

The licence-holder shall appoint persons, who are responsible for the emergency planning, physical protection and nuclear material control. The persons who are appointed to the duties referred to above must have an approval granted by STUK for their specific jobs.

The duties, powers and responsibilities of the responsible manager of a nuclear facility, his deputy and the rest of the personnel needed for the operation of the nuclear facility shall be determined in the Administrative Rules accepted by STUK.

If a person is to be accepted as a responsible manager, he shall have e.g. the following qualifications, unless obviously unnecessary due to the nature of the operations planned:

  • he has a university degree suitable for the job if he will work as the responsible manager of the operation of a nuclear facility;
  • he has the technical expertise in the field of nuclear energy required by the job and, in particular, expertise concerning the safe use of nuclear energy;
  • he has enough practical expertise of the field;
  • he is sufficient familiar with nuclear energy legislation and with regulations issued on the basis of the legislation.

No special examination for a responsible manager is needed but in practice the persons are well known to STUK staff.

Licensing of operators

The licensee has primary responsibility for the development, maintenance and verification of the qualifications of the nuclear power plant shift supervisors and operators and for the conducting of their examinations. The licensee shall, in addition to the examinations, also otherwise follow how the shift supervisors and operators perform their duties.

The licensee shall file an application with STUK for an approval for a nuclear power plant shift supervisor or operator. The application shall concern the shift supervisor or operator duties mentioned in the licensee's Administrative Rules and organisational manual. The operator licensing application shall in addition be focused on the manipulation of the reactor or the turbine, or both. A license granted for a shift supervisor and an operator is plant unit specific and of a specified duration. A licensed shift supervisor can also perform the duties of an operator at the plant unit in question.

Before the licensee files an application for STUK's approval of an applicant as a shift supervisor or an operator, the licensee shall find assurance of the applicant's competence to discharge his responsibilities.

The following are preconditions for the approval of a shift supervisor or an operator:

  • medical examination
  • verification of trustworthiness
  • written examination
  • on the job training
  • personal demonstration of professional skill at a simulator
  • oral examination.

In the application of a shift supervisor's and an operator's (as well as a control room trainee's) license the following information shall be presented:

  • a reference to the previous approval decision
  • basic education and work experience
  • extract from the training register showing initial, refresher and continuing training
  • certificate of medical examination
  • proof of verification of trustworthiness
  • records of the written examination
  • information relating to on the job training and work in the main control room (not applicable to the control room trainee application)
  • record of demonstration of professional skill at a simulator (not applicable to the control room trainee application)
  • record of oral examination (not applicable to the control room trainee application).

At least eight weeks of simulator training must be provided during the initial training period. At the end of simulator training a personal demonstration of professional skill at a simulator shall be arranged for the applicant.

The renewal of a shift supervisor's or an operator's license presupposes

  • medical examination
  • verification of trustworthiness
  • regular participation in main control room duties
  • regular participation in refresher training
  • teamwise demonstration of professional skill at a simulator
  • oral examination.

In the application for the renewal of a shift supervisor's or an operator's license the following information shall be presented:

  • a reference to the previous approval decision
  • basic education and work experience
  • extract from the training register showing initial, refresher and continuing training
  • certificate of medical examination
  • proof of verification of trustworthiness
  • information relating to control room work
  • record of teamwise demonstration of professional skill at a simulator
  • record of the oral examination.

It is mentioned in the Loviisa and Olkiluoto annexes that STUK has some concerns about the availability at the plants of qualified engineers to handle complex safety design issues. Are there any requirements in this respect on in-house staffing as compared to the use of external consultants?

Basic qualification requirements for utility personnel are given in the Nuclear Energy legislation and in the Decision 395/1991, and in more detail in Guide YVL 1.7. The basic approach is that each staff member shall be competent for his/her job.

The general policy of STUK is that if a licensee uses external contractors, the contractor has to fulfil similar requirements as would apply to the licensee if it were carrying out the task by itself. In addition, the licensee is required to confirm its acceptance of the contractor documents when submitting them to STUK.

The background of the concern expressed in the Loviisa and Olkiluoto annexes does not relate to lack of qualifications or competence of individual staff members but rather on the occasional shortcomings in the co-operation and responsibility sharing between various units existing in the licensee organisations (including of-site units such as Fortum Engineering Ltd). The concern was expressed to provide the licensees with guidance for their future actions, both in terms of day-to-day life and possible organisatorial issues. The licensees have already taken the first steps to improve the situation, and the consolidation of this cultural development process in underway.

STUK has not formulated explicit requirements to address this issue because the existing requirements already cover necessary aspects related to individual qualification and organisatorial structure. STUK emphasises that both technical and cultural safety should be continuously improved.

Which is the scope of simulation in the Finnish full scale simulators? Which practice exists with regard to implementation of plant modifications in the simulators?

Loviisa simulator

Loviisa simulator is a plant specific full scope simulator. The simulation of all equipment is about 85–90%. The simulator has a replica process computer. All modifications are carried out immediately after outages e.g. during the same autumn as outages take place and before starting simulator training.

Olkiluoto simulator

The training simulator in Olkiluoto Training Centre is full scope and plant specific according to the Standard ANSI/ANS-3.5's relevant version. Simulator can simulate even beyond design basis accidents with the exception of core melt. Both Olkiluoto units are identical and the reference unit is OL1. Normally applied system concerning plant modifications has been to implement modification first into OL1 and next year to OL2. According to applied principles reference unit modifications, determined to be relevant to the training program have been implemented to the simulator within 24 months of the reference unit in-service date. Certain modifications may also be tested and verified in the simulator before implementing into the plant. These are mainly of type like control room layout modification, ergonomic design modification or signal presentation modification in cases where control room crew comments on various alternatives is needed. Training simulator has not been used as an engineering simulator.

Please provide additional information about financing provisions taking for decommissioning of nuclear installations.

In the Nuclear Energy Act, nuclear waste management is defined so that it includes decommissioning of nuclear facilities. Consequently the financial arrangements for nuclear waste management, referred to in Sections 35 to 53 of the Nuclear Energy Act, apply to decommissioning as well.

The decommissioning costs estimates are based on relatively detailed decommissioning plans that are updated and reviewed by the authorities every five years. Final disposal of dismantling waste is included in these plans. The financial liability is reviewed by the authorities annually.

According to the nuclear legislation, the financial liability for future waste management shall be fully covered by the licensee. They pay annually assets to the State Nuclear Waste Management Fund, which is under the control of the Ministry of Trade and Industry.

The financial rules are such that the waste management fund reaches the full liability not later than 25 years after the commissioning of a NPP. Prior to that, the remaining liability is covered by securities. Currently, the Finnish NPPs have been in operation about 20 years on the average and consequently about 80 % of the future waste management liability is covered by the assets in the State Nuclear Waste Management Fund.

Article 12 – Human Factors

How is safety culture emphasised in operator training programs and training courses of other plant personnel?

The possibilities to indoctrinate safety culture by training are limited because the basis of safety culture is deep in the way the organisation functions as a whole. In any case, there is also formal training on this topic.

Loviisa plant

The Loviisa plant Quality Assurance manual contains a requirement to include training topics and targets for development to the training programmes, to plans of actions and to the yearly QA-evaluation reports.

On the basis of this requirement the safety culture is included as a topic to the training programmes. The operators and also other personnel training is based on individual initial and refreshing training plans. The plans contain the topics, type and duration of training and the schedule for the refreshing training. One topic of these plans is now the safety culture.

In addition, there has been some articles in local plant bulletin and briefing connected to the safety culture. The safety culture is also been improved by developing separate activities such as learning from events.

Annual retraining to plant personnel on foreign (and domestic) nuclear incidents include safety culture & human factors related examples.

Olkiluoto plant

Safety culture is emphasised in most of the courses held to operational and maintenance staff at the TVO plant.

Principles of safety culture and safety culture as a topic are included in the initial training which is mandatory to every incomer concerning TVO's own as well as subcontractors' personnel. The initial training is required before the access card to the plant will be granted.

Safety culture is one of the topics in orientation course, which is mandatory to every new member of TVO staff .

Safety and quality oriented attitude has been determined as one of the five most important areas in TVO's training for year 1999. The determination has been made by company's Training Board in August 1998. Due to this determination:

A lecture on Safety Culture emphasising prevention of Human Errors has been held to operational and maintenance personnel in beginning of the year.
A lecture on safety- and quality oriented attitude has been included in the annual training programme in 1999. The lecture is drawn up for all the TVO's personnel.

What are the policies established at both nuclear sites to encourage plant staff to report “near-miss” events on a voluntary basis?

Loviisa plant

The Loviisa plant has a quality policy which stresses the importance of being alert in all situations. If deviations are found they should be reported to supervisors or to the management. It is clearly understood by everybody that people are not punished in any manner on “near-miss” events.

In training programmes this attitude is encouraged (see other answers above concerning safety culture). Also a regular initiative-program is running at the Loviisa plant.

Olkiluoto plant

For reporting of “near-miss” events in the power production process, a network form has been developed. A separate guide on how to obtain the form base is available.

“Near-miss” events in the process are characterised by a defect or a deficiency having been observed in instructions, working papers etc., which could have disturbed the process, had it not been detected in time, or by an erroneous steering, repairing or an another action having been close to being performed in the process operation or maintenance. Also inappropriate working methods or work arrangements belong to the scope of “near-miss” event reporting.

Reporting of “near-miss” events is important in order to prevent similar events from causing disturbance in circumstances where another abnormal factor would contribute to the development of the disturbance. It is clearly understood by everybody that people are not punished in any manner on "near-miss" events.

“Near-miss” events in the process are dealt with in weekly meetings.

Clear device failures are reported in the job order system as an observation of failure.

“Near-miss” events related to industrial safety have their own reporting form and handling system.

It is reported in the Loviisa and Olkiluoto annexes that root-cause analysis is made of events which are considered as significant. Which are the applied criteria on significance? Which methods are used for root-cause analysis? How are the results of these analyses used by the plant organisations and STUK? Which is the general status of human-factors analysis as compared to technical event analysis? Which human-factor expertise is available for the NPPs and STUK?

Loviisa plant

Root-cause analysis is performed for important operational events whose root causes are not obvious. The specific safety significance could include the following:

  • Some special situations like breach with technical specifications or a fire (a list specified in Guide YVL 1.5).
  • Common-cause failures and recurrent deviations.

The Loviisa plant is using currently as a root cause analysis method its own modification of HPES-method (Human Performance Enhancement System) but also other methods may be used in different cases.

Root-cause analysis reports are distributed widely and they are used in training. The whole internal operating experience programme is assessed periodically.

Human factors are always taken into account in root-cause analysis (and in other event investigations also). At the plant site there is not personnel with behavioural education. Persons in the operation organisation assess human factors with long experience achieved during many years of power plant operation.

Olkiluoto plant

Detailed root-cause analysis is made to events which are considered as very complex or contain significant deficiency in the operations of the organisation and which have not been analysed in detail in the official event reports. Root-cause analysis method (MTO) is developed from INPO´s HPES. The safety committee, plant meeting or operational experience feedback group can decide the need for root-cause analysis. The safety office is responsible to perform the analysis. The results of the analysis are handled as any other event report except the organisational problems which are handled either in the plant meeting or in the safety committee. The number of root-cause analysis is much less than the number of technical reports. The expertise used on human factors has been this far bought from Technical Research Centre of Finland. TVO has also had several courses on MTO-method and expertise gained during the Olkiluoto 1 and 2 PSA work related to human-factors is available.

IAEA's ASSET method is used to analyse operational events. A mission was conducted to analyse 132 operational events occurred during years 1993–1997 at the Olkiluoto plant. The selection of events was based on reporting criteria to the regulator. The events analysed had resulted in special report, scram report, disturbance report or immediate limitation of operation according to the Technical Specifications. The next mission includes the events from years 1997 and 1998. The objective is to use the ASSET method as an every day tool in analysing any deviations discovered in the operation of the plant.


Operational events shall be reported to STUK according to Guide YVL 1.5.

STUK controls the operational experience feedback arrangements of the licensees as part of its own inspection activities. This control includes a review of instructions and reports submitted to STUK. STUK also controls on-site that the instructions are complied with.

STUK assigns its own investigating team to look into operational events deemed to be of special importance. Such an investigation is carried out especially when STUK considers an independent investigation is necessary due to the nature of an event, and also when it is assessed that amendments to requirements or control procedures established by STUK may help prevent the recurrence of an event.

Traditionally event analysis have been more technically oriented but the importance of human factors has been recognised. Inspectors have attended to courses on MTO-method organised by Finnish NPPs. STUK is developing expertise on human factors and at the moment STUK has one person with behavioural science education.

Procedures for anticipated operational occurrences and accidents are in use. The procedures have been verified during operator training on simulators and independently analysed by STUK. But the organisation is not described: is there a safety engineer?

Yes, there is. Both Finnish plants have the similar approach: During normal operation the shift supervisor accompanied by two operators is in charge of the operation in the main control room. In case of unusual events, whatever the situation might be, he can ask for advise from the plant safety engineer, who is available on 24-hours-a-day duty system. The safety engineers on duty are well experienced shift supervisors. If the event is changing to an emergency, the plant emergency organisation as well as the authorities are alarmed according to the plant Emergency Plan. The shift supervisor is acting as the emergency manager until the emergency manager will arrive at the plant and assume the command responsibility. The duties of the emergency manager as well as the organisation are described in Finnish regulations, however, the structure of the organisation is left to be decided by the licensees. STUK evaluates the performance of the plant emergency organisation during exercises. More detailed information is given in Guide YVL 7.4.

Shutdown situations have particular features concerning human factors: are there specific measures (procedures) relating to shutdown situations?

In Finland the risks occurring at shutdown situations have been assessed as part of the plant specific PSAs (see chapters 2.4 of Annexes 5 and 6 of the Finnish report). Also the plant Technical Specifications cover all the shutdown modes. In general, the definition used in all Finnish safety regulations for “normal operational mode” includes i.a. planned outages of the plant. This means that no exemptions from the existing safety level are accepted during the outages.

The detailed requirements for ensuring the safety during outages are given in Guide YVL 1.13. It includes i.a. requirements for the availability of most important safety functions, outage planning, emergency preparedness/physical protection/fire protection during the outage, radiation protection, administrative work controls, quality system functioning, independent safety supervision of the outage works and reporting to the authority. These requirements—and a lot more—have been included into the plant procedures.

The information provided in this article does not clearly stated who deals with human factors to prevent, detect and correct human factors problems. Neither does it state if a particular organisation or task is assigned to a group within the licensee' organisation. Please provide additional information on whether there is a systematic evaluation of human factor performance in situations other than abnormal events: i. e.: as part of improvements due to research in this area.

The responsibility to take into account human factors when using nuclear energy belongs primarily to the licensees as well as the overall responsibility for safety. The nature of human factors is to some extent equal to QA in the sense that it can hardly be taken care by some organisational units but more like as a part of all activities affecting the safety. The Finnish licensees don't have separate "human factors units" with experts having behavioural education. If needed, outside technical support organisations are used for human factors' theoretical analysis. In Finland this expertise is available by the Technical Research Centre of Finland/Laboratory for Industrial Automation/ Unit for man-machine psychology (6 employees at the moment). This unit has carried out research work at NPPs e.g. in the area of NDT–realibility and operational events analysing procedures. The authority STUK has at the moment one employee having behavioural education (social psychologist) and one senior technical expert specialised in human factors area. In general, e.g. the operational experience feed-back analysis (root cause analysis) are managed by well experienced senior experts having engineering background both in the licensees' organisations and in STUK.

Article 13 – Quality Assurance

It is not clear from the report if Loviisa has a supplier audit programme. In what extent do the two Finnish operators cooperate in the auditing of suppliers?

It is a general requirement in the YVL Guides that the licensee must describe the QA programme of each supplier of vital safety related equipment, and that quality audits at factories must be conducted during and after the fabrication.

At the Loviisa plant the supplier control group is responsible to control and supervise the activities connected to the evaluation and control of suppliers. The participants of the group are from separate organisation units at the plant, and the chairman is maintenance manager. This group compiles the yearly supplier audit plan on the basis of information given by its members. This yearly audit plan is regularly updated when new information exists for example from implemented audits. The project manager is also responsible to compile the project specific supplier evaluation program. This program is approved by the supplier control group.

There has been some co-operation between the Olkiluoto and Loviisa plants but it is not a regular activity.

It is mentioned that QA- programmes of the licensees and of the main suppliers are subject to approval by STUK? What is the reason for this provision? Does STUK also approve all modifications to these programmes?

According to Guide YVL 1.4, the Quality Assurance programmes of the licensees are subject to approval by STUK. STUK approves also all modifications to these QA-programmes.

As regards Quality Assurance programmes of main suppliers, the detailed requirements on the information to be provided are given in separate YVL Guides. The adequacy of Quality Assurance programmes are reviewed e.g. when evaluating the competence of pressure vessel manufacturers. STUK does not grant a separate formal approval stamp to suppliers or their QA programmes, but only an approval to supply specific equipment or services to the licensee in question.

It is considered that this approach does not mean any transfer of the responsibility from the licensee to the regulatory body. It is quite opposite because a detailed review by STUK requires the utility to make a proper assessment of it suppliers before they can present their case to STUK.

Could Finland indicate if the regulations are in accordance with international recommendations from ISO and IAEA?

Requirements for Quality Assurance have been specifically set in Guide YVL 1.4, YVL 1.9 and YVL 6.7. These Guides have been prepared taking specifically into account the IAEA safety series documents for QA, the documents being referred in the Guides. Guides YVL 1.4 and YVL 1.9 have been scheduled to be revised this year.

Please describe what further developments are needed in the existing quality system of the Loviisa and Olkiluoto NPPs, which the report mentions as a result from the review based on the Decision 395/91.

The reported conclusion was that “...Quality System meets the provisions set forth in Decision 395/1991”. However, relating to the review of TVO, STUK wanted to highlight the importance of continuous development of quality by mentioning some findings of the reviews carried out by STUK staff and a quality expert contracted by STUK. These findings have been included in the long-term development program of TVO.

An observation worth mentioning was that the formal QA system at the Olkiluoto plant was still based on the old Appendix B of 10 CFR 50 in the U.S. regulations, while the management approach had followed the general developments in modern industries. Thus there were like two parallel approaches to high quality operations, and there was a need to streamline some activities.

As regards the Loviisa plant, its was e.g. noted that the quality system does not include proper requirements on the design related to the implementation of safety principles. Extension of the quality system to also cover this area is a subject for development in the future.

The issues related to Quality Assurance are discussed in more detail in Annexes 5 and 6 of the Finnish report.

Article 14 – Assessment and Verification of Safety

Does maintenance during shutdown result in special safety precautions?

Yes, it does. This issue is discussed under Article 12.

The risks related to outage situations are covered by deterministic and probabilistic safety analyses, which are discussed in more detail in Annexes 5 and 6 of the Finnish report.

Does Section 6 of the Decision 395/1991 regulate the frequency of updates of the accident analyses and PSAs. If not, where and how is this requirement set?

There is no formal requirement on the update frequency, but in practice some updating of the analysis is done all the time because there are people working full-time on safety analysis both with the licensee and the regulatory organisations. Comprehensive assessments of the validity of analysis have been done about every 10 years, in connection with the re-licensing of the operation of the Finnish plants. In fact, the most recent re-licensing involved repeating of almost all previous analysis, using state-of-the-art codes and data.

Article 15 – Radiation Protection

What are the average occupational doses and the maximum exposures values for the personnel at the nuclear power plants and what are the collective doses for each nuclear power plants or per MWh produced?

The distribution of individual doses in Loviisa and Olkiluoto NPPs in 1998 can be seen in Table 1. The average dose per worker in 1998 was 1,8 mSv.

As regards collective doses, information is given in Annex 6 of this report.

Table 1. Number of workers in NPPs and the dose distribution in 1998
Dose (mSv) Loviisa Olkiluoto
under 0,5 248 607
0,5–1 132 273
1–2 119 284
2–3 72 136
3–4 47 90
4–5 47 42
5–6 34 25
6–7 25 19
7–8 14 11
8–9 12 5
9–10 12 3
10–11 10 3
11–12 9 2
12–13 2 2
13–14 5 1
14–15 5
15–16 1
16–17 1
17–18 4
18–19 1
19–20 2
over 25

The individual and collective doses as well as releases of radioactive materials from the plants are discussed in more detail in Annexes 5 and 6 of the Finnish report.

How is the ALARA principle being handled in practice and is it a living ALARA? What was the success of the major measures or efforts to reduce the exposure?

The Loviisa and Olkiluoto plants have established their own ALARA programmes, which are up-dated yearly. The requirements for the implementation of ALARA principle are stated in YVL Guides (especially YVL 7.9). STUK inspects annually the implementation of the programmes in periodic inspections to the plants.

At the Olkiluoto plant the programme mainly focuses to operative radiation protection procedures and to water chemistry. One of the main achievements of the programme has been the replacement of components containing stellite, which work is still continuated. Fixed shields have been installed to protect workers during outages. There have also been improvements in purification systems of primary coolant.

At the Loviisa plant one of the most significant actions related to the ALARA programme was the full decontamination of Loviisa 2 primary circuit in 1994. It has been estimated that the dose saving in the years to follow the decontamination has been at least 8 manSv.

The Loviisa and Olkiluoto plants have also focused to work-planning and training in their ALARA programmes.

Even though there were large modernisation programmes, including system modifications at both plants during recent years (starting 1996), no major increase in collective doses could be seen (see Annex 6). In that respect the ALARA programmes have been successful.

What were actual radioactivity release values and how did they develop over the past decade?

The general trend has been decreasing during the past decade. This can be seen in the estimated doses for an individual of the most exposed population groups in the vicinities of the Finnish plants (see p. 20, fig. 10 in Annex 9 of the Finnish report). This trend is mainly due to the improvements made to the liquid waste handling systems before and during the modernisation projects at the Loviisa and Olkiluoto plants.

The major part of estimated doses has been due to liquid releases (Co-60 being the dominating nuclide). However, after the improvements made at the Loviisa plant in the beginning of 1990's, the share of liquid releases declined sharply, and the major part of the estimated dose is usually due to airborne releases.

Could Finland indicate the individual and collective effective doses and the releases in the environment?

This issue is discussed above.

The report does not clearly stated which is the role of STUK (if any) in the environmental monitoring program.

According to Guide YVL 7.7 the licensee is responsible for the monitoring of radioactivity in the environment of the plant. The monitoring program has to be accepted by the Nuclear Reactor Regulation department of STUK. So far both licensees have ordered the environmental monitoring from the Research and Environmental Surveillance department of STUK. It has to be kept clear that only the first department mentioned is responsible for the regulation, and the second department is involved only because it can provide competitive services to the licensee. The first department audits the second one as it would audit any external research organization. Besides, the measuring and analysis methods of the second department are just now being formally accredited by the national accreditation organization.

Please provide additional information on the periodicity of dose estimations to population due to normal plant releases and the methodology used for this purpose.

The doses are annually estimated for an individual of the most exposed population group in the vicinities of the Finnish plants. The estimation is based on actual measured airborne and liquid releases and actual meteorological data observed on-site. The methods used in the estimation take into account all the relevant pathways. The over-all methodology is highly conservative.

The requirements for the dispersion and dose estimations are provided in Guides YVL 7.2 and 7.3.

With respect to the 5 mSv/year for population limit stated in Section 11 of the Decision 395/91 (postulated accidents). The reviewer found that Section 28 states a 100 mSv/year limit for Nuclear Installations with an operations license granted prior the Decision entered into force.

The limit 5 mSv for postulated accidents is especially intended for new plants and would require modifications to the current PWR designs (the limit would not allow direct release from steam generator safety valves to the atmosphere, which is the situation in all current PWR's of the world).

According to the accident analyses, a major leak from the primary to secondary circuit at the Loviisa plant could result in an essential radiation dose exceeding 5 mSv to an individual of the population. The methods used in the calculation are very conservative.

The licensee has paid special attention to an accident situation where a large reactor coolant leak from the primary circuit to the secondary circuit could occur as a result of a steam generator inner damage (see Annex 5 of the Finnish report). Since 1994, the licensee has implemented modifications intended to prevent the fuel damages and large releases of radioactive materials in connection with these kind of leaks. The main purpose of the accident management measures has not been to limit the release of the postulated accident and the theoretical radiation dose of an individual in the surroundings, but to prevent a severe reactor damage during such an accident.

Please provide additional information on fundamentals of rules and regulations to limit radioactive releases to environment. How are the ALARA principles applied?

Decision 395/1991 sets the following: “The limit for the dose commitment of the individual of the population, arising from normal operation of a nuclear power plant in any period of one year, is 0.1 mSv. Based on this limit, release limits for radioactive materials during the normal operation of a nuclear power plant are to be defined.” More detailed requirements are given in Guide YVL 7.1.

The release limits are calculated based on the dose constraint 0.1 mSv/year. The licensee is responsible to determine the release limits, which are subject to the approval of STUK. The dose constraint is for a site, and it includes all on-site activities.

It is required that releases are as low as reasonably possible. In practice this means low releases if compared internationally, and the use of adequate methods and instruments. The releases have annually been less than 5 % of the limit, but major efforts have been done to reduce the releases even further. These efforts have brought good results.


Article 16 – Emergency Preparedness


In assessing the radiological release in case of a nuclear accident, have “worst case scenario“ assumptions been applied?

The emergency plans are based on release fractions of most significant isotopes that are in the same range as in Chernobyl, but releases are assumed in much shorter time and at worst elevation.

Decision 397/1991 provides in Section 3 the following:

Emergency planning shall be based on the analysis of nuclear power plant behaviour in emergencies and on the analysis of the consequences of emergencies.

Guide YVL 7.4, Section 2.2, includes the following requirements:

The progress and consequences of typical accident scenarios shall be analysed to facilitate emergency response and the classification of emergency situations. Highly unlikely accidents, which may have severe consequences, shall also be analysed.

Who (which positions) in the Finnish system are primarily responsible for on-site accident management measures and off- site protective measures respectively?

Rescue operations management is based on the Fire and Rescue Services Act (599/75) and Decree (1089/75). The rescue management system is based on the allocation of responsibilities at local level in Finland. The Ministry of the Interior in co-operation with the central government authorities is responsible for measures at national level. Corresponding responsibility at provincial level rests with provincial governments, which act in co-operation with other local authorities. Municipal chief fire officers and rescue authorities manage actions in their fields of responsibility. By virtue of Section 29 of the Fire and Rescue Services Act (559/75), the Ministry of the Interior can order an official in its employment to take command of the rescue operations.

In plans prepared for nuclear power plant accidents, it is assumed that when rescue operations are initiated, the municipal chief fire officer of the area of location of the power plant takes charge of the rescue management and initiates the necessary notification, alerting and rescue operations. In a severe accident, which requires extensive action, the aforementioned national command system is taken into use to the extent required.

On-site emergency manager, who is the head of the emergency organisation at the nuclear power plant, is in charge of the on-site emergency response and liaison with authorities (Decision 397/1991). He is obliged to comply with the instructions for the rescue operations issued by the authority who has command responsibility for the rescue operations. In practise, however, the licensee is always responsible for nuclear safety and radiation protection on-site. On-site emergency manager is also responsible for the off-site rescue operations and countermeasures until the responsible authority takes charge and announces about it.

It is mentioned in the report that findings from an inspection by STUK pointed at some improvement needs, for instance in the classification of emergency situations and arrangements for accident management. Which improvements are planned?

The Report mentions that STUK performed two on site emergency preparedness exercises in 1997 and 1998. Weaknesses such as protection of personnel during an accident, classification of emergency situations, dispositions on handling accidents and information to the public in advance, were defined. How does STUK define whether these weaknesses are resolved? Please explain...

Based on a comprehensive re-evaluation of the emergency preparedness plans at the Finnish nuclear power plants in 1997–1998 STUK has required licensees to develop and improve certain procedures. STUK has assessed this development work during periodical inspections and in other communications with the licensees, and through the review of up-dated emergency plans.

Some examples of this development are described below:

A third level in the classification of emergencies “emergency alert” has now been included into the emergency plans of the nuclear power plants. This level means an elevated risk of a severe accident. The first level is “general emergency” and the second “site emergency”.

Advance information to the public about emergency measures has been delivered. Protection of the personnel during an accident has been improved based on analyses related to the radiation situation outside and inside the power plants during severe accidents and on evaluations concerning protecting measures.

Alarm methods have been enhanced in the residential area for outage personnel in the vicinity of nuclear power plants. The evacuation actions of the staff from the sites have been analysed.

Arrangements for accident management have been developed (such as equipment for the communication between authorities and the command centres at the nuclear power plants and emergency support facilities). Additional guidance for keeping logbooks and making situation assessment has been given.

A general description of the Guide YVL 7.4 should be given in order to have a better understanding on emergency preparedness. In particular additional information on the following points would be appreciated :


  • classification of emergencies,
  • general national organisation,
  • description of on-site and off-site emergency plans including supporting organisations,
  • criteria and guidelines used by the operating team to take decision about entering into emergency situations,
  • measures taken to protect the public (sheltering, stable iodine tablet distribution, evacuation), criteria and guidelines used to decide if measures are needed, lessons drawn from on-site and off-site emergency exercises.


General description of the YVL 7.4

The general regulations for the emergency response arrangements at the nuclear power plants are provided in Decision 397/1991. The detailed requirements are given in Guide YVL 7.4. Such requirements relate e.g. to emergency planning and organisation, management of an emergency situation, worker safety and radiation protection, radiation measurement and to post emergency measures. The guide also includes requirements for the information to the public, protection of the public and for maintenance of preparedness. Also regulatory control activities are described.

Classification of emergencies

According to Guide YVL 7.4 emergencies shall be rated as “site emergency” and “general emergency” based on their severity and manageability. During site emergency the nuclear power plant's safety deteriorates or is in danger to deteriorate significantly. During general emergency there is a threat that radioactive materials will leak in the environment, which may require protective measures in the vicinity of the plant.

Preparedness situations also include “emergency alert”, which involves alerting the nuclear power plant emergency organisation to the extent necessary to ensure the plant safety.

General national organisation

The national level consists of organisations such as the Ministry of the Interior, the Ministry of Social Affairs and Health, the Information Unit of the Council of State and the Ministry of Agriculture and Forestry etc. as well as the advisor organisations STUK and Finnish Meteorological Institute and when information to the public is considered the Finnish Broadcasting Company. Second level is the regional administration (provinces) and third is the municipal level. In plans prepared for nuclear power plant accidents, it is assumed that when rescue operations are initiated, the municipal chief fire officer of the area of location of the power plant takes charge of the rescue management and initiates the necessary notification, alerting and rescue operations. In a severe accident, which requires extensive action, the aforementioned national command system is taken into use to the extent required.

On-site and off-site emergency planning

On-site emergency plans are prepared by the nuclear power plant and approved by STUK. The emergency plan shall cover the following: classification of emergency situations, descriptions of organisation and arrangements for alerting and communication, management of an emergency situation and assessments of the situation, worker safety and radiation protection and on- and off-site radiation measurements during an emergency situation. The plan shall also include provisions for information, description of rooms, equipment and facilities for emergency situations, post emergency measures and a description of the maintenance of preparedness.

Ministry of Interior is the leading organisation giving the directives and general commands for off-site emergency planning. Regional administration (provinces) supervises the local off-site rescue organisations, which prepare special plans for nuclear power plant emergencies. The Ministry of the Interior, STUK and other authorities review these emergency plans. The plans include the command plan, the agenda of the command centre, the plan for public information etc.

The national level organisations such as the Ministry of the Interior, Ministry of Social Affairs and Health and Ministry of Agriculture make their own plans, as well as the advisor organisations STUK and Finnish Meteorological Institute, and also Finnish Broadcasting Company.

Criteria and guidelines for decision making by the operation team about entering into emergency situation

YVL 7.4 states that the progress and consequences of typical accident scenarios shall be analysed to facilitate emergency response and the classification of emergency situation. The emergency plans of the nuclear power plants are generally based on this requirement. The control room personnel identifies the emergency situations on the basis of their technical instructions.

Criteria and guidelines for measures taken to protect the public

In Finland criteria for measures taken to protect the public are based on the recommendations of IAEA (Intervention Criteria in a Nuclear or Radiation Emergency, Safety Series No. 109, IAEA, Vienna, 1994). STUK has also given guidance for operational intervention for short-term countermeasures (Table). These levels are based on external dose rates and used when the composition of radioactive plume is not known.

Protective action/ STUK Operational intervention level as an external dose rate
Sheltering 100 µSv/h
Iodine prophylaxis 100 µSv/h
Preventive relocation 100 µSv/h
Temporary relocation 100 µSv/h
Permanent relocation 500 µSv/h

Lessons drawn from on-site and off-site emergency exercises

The on-site emergency organisations have developed their emergency facilities, log book keeping and alerting approach (the first alert from nuclear power plant to the authorities is sent by telefax and phone to the expert on-duty at STUK) based on lessons drawn from joint on-site and off-site emergency exercises. An important topic for the on-site emergency organisations in every exercise is to maintain and develop the communication with and information to STUK and the local rescue authorities.

The off-site organisations have strongly developed their alerting approaches and actual communications, and also provision of information to the general public and to the international community.


Article 17 – Siting


Environmental Impact Assessment (EIA) procedures have been applied in practice, e.g. when modernising and increasing the power levels. What are the differences in the scope between the EIA for modification and the EIA presented with the application for a Construction License?

Environmental Impact Assessment procedures have been defined in Finland in the Act (468/1994) and in the Decree (792/1994). Environmental Impact Assessment procedures are applied to projects, which are listed in Section 5 of the Decree. The list includes e.g. nuclear power plants and their radioactive waste treatment, storage and disposal facilities. Procedures are also applied to significant modifications of the projects in question, if the modification has significant environmental effects comparable to the effects of the projects listed in Section 5.

As regards new nuclear facilities, the report on the Environmental Impact Assessment shall be included in the application for a Decision in Principle. This is a political decision on a new project, and must be made before an application for construction license can be filed. An Environmental Impact Assessment for a modification is carried out before the modification is implemented.

For an Environmental Impact Assessment a programme shall be established and submitted to the competent ministry, which is the Ministry of Trade and Industry in the case of a nuclear facility. The ministry then gives its statement on the programme, and it may require modifications to the programme. In the case of a nuclear facility the programme is evaluated also by STUK. After the statement of the ministry the programme can be implemented.

The contents of the programme and the report are generally described in the Decree (792/1994), but there are no specific requirements for the case of a modification. As regards a new nuclear facility the programme shall be comprehensive covering all the environmental effects. Respectively, the programme on an Environmental Impact Assessment for a modification may emphasise the effects of the modification in question. However, the programme is evaluated by the authorities on a case by case basis.

Have state-of-the-art seismic analyses been performed for the Loviisa and Olkiluoto nuclear power plant sites (e.g. probabilistic seismic hazard analyses, seismic margin analyses, seismic PSA)? Did these analyses result in any upgrading? If such analyses are planned, what is the schedule for their completion?

Which seismic characteristics (peak ground acceleration level, intensity, response spectra etc.) have been assumed for the Loviisa and the Olkiluoto sites, based upon which standards and calculations?

Loviisa plant

No seismic criteria or analysis was used during the design stage of the Loviisa plant, based on the well known low seismicity and because design rules for normal (power plant) construction were considered to account for credible seismic loads also. However, seismic analysis have been performed later on when the plant was already in operation.

Seismic hazard curves for Loviisa site (annual exceedance probability as a function of peak ground acceleration) have been determined based on the Fennoscandian seismic records (magnitudes, locations, times), integrating over the source area with a relevant attenuation function. Seismic response spectra of important structures of the plant have been calculated, and fragility curves of plant components have been determined. Lowest capacities were found in large water tanks.

All this information has been used to calculate which initiating events (of the PSA model), and at which rates, are caused by seismic events. Seismicity in Finland is so low that even with the conservative assumption that each initiating event leads to core damage, the median CDF is about 4.4E-7 per year. (The mean value 3.6E-6/a). Thus, no plant modifications were necessary based on the seismic risk assessment.

Olkiluoto plant

Seismic probabilistic risk assessment (SPRA) has been performed for the Olkiluoto plant. Seismic hazard at the Olkiluoto site is low. Located in a relatively low seismic region, with the 100,000 year event peak ground acceleration of less than 0.1 g, the plant was not designed for the seismic loads. As a hard rock site most of the energy content of the ground motion is at relatively high frequencies and consequently ground motion displacement amplitudes are small.

The PRA was generally similar to the seismic PRAs conducted for several US power plants as part of the Individual Plant Evaluation for External Events, consisting of seismic hazard assessment, walkdown, system analysis, structure response calculations, component fragility analysis and risk quantification. Guidance from IAEA, USNRC, and EPRI documents was followed to identify seismic initiators, plant safety functions, and associated plant equipment critical to seismic safety. A total of approximately 800 components were placed on the seismic equipment list. Sensitivity and uncertainty studies were also performed as part of the quantification process.

The total combined core damage frequency due to all the accident sequences caused by an earthquake is nowadays approximately 5E-6 p.a. This is based on the assumption that the anchorage of batteries and inverters has been upgraded. TVO has already completed a significant number of these upgrades. The main control room lighting fixtures and ceiling grid will be upgraded by the end of the year 1999.

In decision 395/1991, it is indicated that the combined effects of accident conditions induced by internal causes and simultaneous natural phenomena shall be taken into account to the extent estimated possible. Could Finland indicate what events have been taken into account?

Loviisa plant

When doing PSA for external events and natural phenomena [floods, fires, extreme weather phenomena (storms, snow, seaweed, fragile ice, low/high temperatures and combinations of these) and seismic events], all internal "random" failure probabilities were also included for all components modelled in the PSA. Thus, the risk was not estimated for external events or internal events in isolation.

Olkiluoto plant

A comprehensive PSA has been performed for the natural phenomena induced initiating events for the Olkiluoto units 1 and 2. A multiphase screening method was used when analysing the potential effects of natural phenomena exceeding the design basis of components, systems and structures.

In the first screening phase about 50 different natural phenomena were included in the analysis. All such phenomena were rejected that could not cause an initiating event (scram or requirement to shut down) and simultaneously exceed the design basis or otherwise degrade the safety systems or functions, leading to core damage with larger than 10-8/year frequency. The frequency estimation was very coarse and conservative in the first phase.

Special consideration was put to the combination of natural phenomena having impact on several systems (High wind may destroy external grid only, but combined with snow fall it may also cause harm to diesel generators' intake air filters. High wind combined with cold weather in early winter, when the sea does not yet have the ice cover, may cause frazil ice degrading the ultimate heat sink and diesel generator cooling simultaneously with loss of off site power.)

In the second phase the potential natural initiators were analysed physically with more exact consequences to the components, systems and structures, and the frequencies were estimated for each initiating event. A new screening was made, and the following phenomena were included in the living PSA model:


  • Seven sea water impurities related initiating events with loss of main feed water and different combinations of safety systems
  • One high wind related initiating event with total loss of off site power
  • Three frazil ice related initiating events with loss of main condenser and different combinations of safety systems
  • Three hoar frost related initiators causing loss of room cooling in reactor building, some isolations and requirement for shut down the reactor
  • Ten lightning related initiating events with loss of main feed water and different combinations of safety systems, e.g. diesel generators

Seismic events, internal fires and internal flooding were analysed separately from the analysis of other natural phenomena.

Article 18 – Design and Construction


Which safety analyses have been made to address the process of design changes and facility upgrading measures? Was there a different scope defined when compared to the commissioning analyses?

The analyses concerning the operation of the Olkiluoto plant have been completely renewed during the modernisation project. In addition to the uprated power level, the analyses have taken into account e.g. the changed reactor power/flow rate area and the structural modifications of fuel rods and reactor internals. To ensure the scope of analyses is adequate TVO has conducted a failure consequence analysis, where the consequences of transients caused by system and equipment failures have been considered from the standpoint of plant operation. Based on the review, the available transient and accident analyses cover well the transients caused by system and equipment failures that may occur in operation.

Fortum Power and Heat Oy has revised in connection with the modernisation of the Loviisa 1 and 2 units all transient and accident analyses included the Final Safety Analysis Report, taking into account the increase of reactor power.

The scope of the revised analysis of both Olkiluoto and Loviisa plants is generally the same as in the phase of commissioning. Due to changed regulatory requirements also some new analyses have been included, e.g. ATWS analysis. It should also be noted that the analyses were made with advanced methods that represent the current state-of-the-art.

Is there an in-service monitoring program implemented to register and evaluate the operational loads on safety related components for comparison with the design basis assumptions?

At the Loviisa and Olkiluoto plants the in-service monitoring of operational loads of safety related components has been carried out since the start-up of the plant operation. The monitoring of loadings has mainly been based on book-keeping of transient operating conditions, such as plant shut-downs, reactor trips, pump start-ups and trips. For each type of transient the associated pressure and temperature fluctuations as well as the expected number of events during 40 years in Olkiluoto and 30 years in Loviisa have been accounted for in the component design specifications. The transients are mostly detected by means of original process instrumentation and data is stored into the transient book-keeping file. This information is checked annually under STUK´s supervision.

At the Olkiluoto plant the continuous monitoring of feed water temperature fluctuations in all four feed water lines is carried out. The results can be seen from a plotter in the main control room and are stored for later evaluation. In addition, there are seven thermocouples on the outer surface of the feed water pipes near the RPV nozzles of two lines in both units.

At the Loviisa 1 thermal fluctuations due to mixing and/or stratification phenomena in two feed water lines near the steam generator nozzle (four thermocouples and two strain gauges per nozzle), in the pressurizer lower nozzle, in the spray line nozzle and in the pipeline between the pressurizer and the primary circuit have also been registered and analysed during about ten years. In 1998 a new measurement system was installed at Loviisa 1 with 155 thermocouples and 6 strain gauges. Currently totally 28 known or potential stratification locations are monitored on-line.

Some of the most important components such as the main circulation pumps and turbines are also provided with an on-line vibration monitoring equipment. These monitoring systems have continuously been extended and upgraded at both plants.

In-service monitoring is also described in Chapters 4.4 and 5.2 of Annexes 5 and 6 of the Finnish report.

It is mentioned that the development of detailed safety requirements and procedures to ensure adequate reliability of digital I & C systems is still underway. How are these issues dealt with today? Which is the scope of the new requirements?

The digital technologies are implemented at the NPPs of Finland in some monitoring and control systems. However, the work on their reliability analysis and the influence on the safety is not completed. How the licence of the Regulatory Body on the commission of indicated technologies was received?

It is indicated that digital I & C is already implemented in some plants and that the development of requirements is still underway. Could Finland explain on what basis the modifications have been approved?

There are several YVL Guides, which deal with I&C systems in respect of the design and operation of the nuclear power plants. Guide YVL 1.0 provides the general design criteria, including criteria for I&C systems. Detailed requirements concerning the transient and accident analyses are presented in Guide YVL 2.2, and the requirements for the reliability and risk analyses are given in Guide YVL 2.8. Guide YVL 5.5 gives the requirements specific to I&C systems. This guide is under major revision in order to provide clear guidance for the licensing of the digital I&C systems. A draft guide has been available already for years, and it has been tested and improved in parallel with the installation of the first I&C systems of digital type. In principle the existing Guide YVL 5.5 is valid but it does not give specific guidance on digital I&C systems.

Some digital I&C systems have already been commissioned for monitoring and control purposes and also for some safety tasks. However, only the safety functions with analogous hardwire backup are credited in the safety assessment, as described e.g. in the chapter 4.1 of Annex 6 of the Finnish report.

The revised Guide YVL 5.5 will cover the design, implementation, commissioning and operation of I&C systems and components at nuclear facilities. The requirements include the design basis requirements and general design requirements, as well as specific requirements for I&C systems and components design, implementation, commissioning and operation. Also the requirements on the monitoring and control of component ageing are given. Technology related special requirements of the digital I&C systems are presented.

Decision 395/1991 also gives requirements concerning the possibilities to avoid, detect and compensate for human errors. The implementation of these requirements is not described: for example, are there design features allowing easy access to the equipment for maintenance or repair?

More detailed requirements on design features aimed to avoid human errors are given in many YVL Guides. For example, easy access was emphasised already in YVL Guides providing general design criteria, in the early 1980's. In the design of plant structures, systems and components the possibility of human errors is taken into account through several design features e.g. allowing for easy access to the equipment, labelling the components with systematic way, using clear colour codes for certain systems, using self-diagnostic systems, fail-safe systems, alarms indicating malfunctioning etc.

The main approaches to reduce the risks of human errors at the operating plants have been the following: evaluation of human errors and their safety significance as part of the PSAs (to recognise the most risky errors), revising plant procedures based on the PSA findings (to avoid or compensate for the errors), analysing the operational events by root cause methods (to prevent recurring events), testing/inspecting and auditing (to detect the errors), installing automated protection systems (to avoid errors or just give more time to the operators to make a diagnosis of the situation) and training of the staff.

Some detailed information can be found in chapters 4.7 of Annexes 5 and 6 of the Finnish report, and in Guide YVL 1.11.

Please provide additional information related to Section 3.1, point 2, expectations to reduce the consequences of accident sequences involving the bypass of containment at Loviisa NPP.

On the basis of PSA-studies it was identified that there is need to reduce the by-pass sequences of both design basis accidents and severe accidents. That is why a considerable effort has been put to reduce these sequences with plant modifications.

One of the main contributors in PSA level 1 core damage frequency is a large primary to secondary side leak, which includes also containment by-pass via the steam line of the leaking steam generator. The plant modifications implemented for this event include a protection system to isolate the steam and feedwater of the leaking steam line, an additional storage of emergency core coolant and improved reliability and efficiency of the spray system of the pressurizer. With these systems the reactor can be taken safely to a cold shutdown state even if the safety valve of the leaking steam generator would stick open at the beginning of the primary to secondary leak. The other system which was identified to need improvements to prevent containment by-pass was the injection water system of the seals of main coolant pumps. It was equipped with new protection signals to isolate leaks outside the containment.

The measures to cope with severe accidents will include means to verify the closing of continent isolation valves in the case of an accident.

After many plant improvements and modifications related to other sequences, the relative risk contribution of the containment by-pass sequences has become larger than before these improvements. Other reasons for this are the overly conservative assumptions used in the modelling in PSA. The utility has started a project where it will reassess the modelling of containment by-pass sequences in order to define the real risk importance more accurately. The need for necessary additional improvements to reduce by-pass sequences will be assessed after this study has been completed.

It is not clear to the reviewer, which is the criteria used to eliminate the consideration of an accident in the design of Finnish NPPs. Could you please explain in further detail?

All events which are considered possible and can be coped with should be taken into account in the design. Some events such as the catastrophic rupture of the reactor pressure vessel do not fall into this category and be excluded from the design basis.

The original dimensioning basis of the containment was a double-ended break of the main coolant pipe or a main steam line break. After the first start-up, the containment system of the existing Finnish nuclear power plants have been upgraded to limit the radioactivity releases of severe accidents. Some work is still underway, such as installation of catalytic hydrogen recombiners and plant modifications which ensure external reactor vessel cooling in connection with any severe core damage.

Severe accidents and ATWS events must be included in the design basis of the future plants according to the current nuclear safety regulations in Finland.


Article 19 – Operation


What measures are planned with respect to the implementation of risk-informed, performance-based regulations?


In Finland, plant-specific level-l and level-2 PSA studies are required by STUK. Plant specific living PSA's have been completed for all operating Finnish plants, including internal initiators, fires, flooding, harsh weather conditions, seismic events for operation mode and internal events for low power mode. These PSA studies are used in support of regulatory decision making and safety management at STUK and at the utilities, respectively.

The guidelines for applying the Living PSA are set forth in Guide YVL 2.8. The Living PSA is formally integrated in the licensing procedure already in the early design and it is to run through the construction and operation phases all through the plant service time.

In compliance with the requirements posed in Guide YVL 2.8 (published 1987 and revised 1997) the licensee has to use the results of PSA in support of safety management in terms of:


  • plant changes and backfits
  • training of plant personnel
  • working out of emergency operation procedures
  • applications of Technical Specifications
  • case by case assessments of risks resulted from component failures
  • risk follow-up of licensee events
  • directing and weighting in- service inspections and testing
  • maintenance and surveillance programme planning
  • new plant designs


Use of PSA for analysis of plant modification and operational events

PSA has been applied much in evaluation of plant modifications. As a matter of fact it is a regulatory requirement that the utilities provide STUK with the assessment of safety significance of the modification in conjunction with the related pre-inspection documentation. The assessment has to be submitted to STUK independent from the safety class which the modified systems belong to. Up to now a number of plant modifications have been performed based on the insights from level 1 PSAs, assigning highest priority to modifications with the most risk impact.

In the area of operational events, PSA is a standard tool to assess the safety significance of component failures and incidents. Accordingly, systematic risk follow-up studies are being made at STUK.

Use of PSA for analysing Technical Specifications

PSA is used to give arguments for temporary exemptions from Technical Specifications. In such a case however it is provided that exceeding of the allowed outage time contributes only a tiny increment to the core damage probability compared with the normal operation.

Furthermore, the meaningfulness of some AOT's given in Technical Specifications has been re-evaluated by PSA techniques. Additional items have been included in Technical Specifications for shutdown states, based on the results from shutdown and low power mode PSA. For example, STUK set forth a new requirement to keep the lower air lock of the containment closed during the maintenance of main circulation pumps at the Olkiluoto plant, because this task contributes to increase the probability of large bottom LOCA of the reactor vessel.

Use of PSA for preventive maintenance

The majority of preventive maintenance work is performed during the annual outage. However, STUK allows preventive maintenance also during power operation if deterministic criteria are fulfilled (e.g. single failure criteria...). PSA is used to minimise the risk deriving from the on-line preventive maintenance.

At the Olkiluoto plant it is allowed to take one redundancy out of service at a time. In 1989, the risk contribution of the on-line preventive maintenance to the mean core damage frequency was approx. 5 %. Later on the maintenance schedule was modified based on insights from PSA, and currently the risk contribution is approximately 1 %.

Use of PSA for ISI/IST

A project dealing with PSA support to in-service inspection (ISI) and in-service testing (IST) is in progress. The aim of the project is to explore on how the plant specific PSAs can best be used effecting ISI and IST. The study is aimed to consolidate the use of PSA in the respective regulatory process. The study is also to make the ISI and IST more safety -effective and also reduce the unnecessary burden caused by too many inspections.

When modernising the plants, up-ratings, major modifications or renovations, which commissioning programme is requested as compared to the requirements for initial authorisation?

After modifications of separate systems, it is generally required that the system must be tested at the same extent as the original respective system. The test programme is submitted to STUK as part of the modification plan, and is approved together with the plan.

Loviisa test operation

Fortum Power and Heat Oy planned and carried out a trial test program, by which it verified the effects of the power uprate on the functioning of the systems and components of the plant. Normal operation and to a limited extent also transient behaviour of the plant were studied. Studies made by means of the plant simulator and the results of transients analyses were used in the planning of the test program. Due to the small number of plant modifications required for the power increase of the Loviisa plant, a relative concise test program was considered as appropriate and acceptable by STUK.

The long term trial test operation of both units was carried out at various reactor power levels, increasing stepwise the power. Transient tests were carried out at two power levels at both units. They were selected so that by means of tests the acceptability of the functioning of the most important process and control systems of the primary and secondary circuit was verified, the number of the tests being as small as possible.

Olkiluoto test operation

Test operation carried out after the recent plant modernisation and power upgrade included the system related tests, plant unit related transient tests and so-called long-term test operations, during which the reactor was operated at an uprated constant power for a longer period of time. Test operations were conducted in stages at different power levels under STUK`s supervision and within the frames accepted by STUK.

Test operation programmes were based on the original test programmes that were ran through during the first commissioning of the units. The programmes were modified taking into account the requirements caused by the modernised systems. One principle was also to minimise the loads to structures and equipment caused by test operation, due to which the different transient tests concerning the behaviour of the entire plant units were evenly distributed, when possible, to both plant units.

Please clarify in more detail the structure, scope and practice of the plant incident reporting programmes.

Incident reporting from utilities to STUK

Reporting requirements are set in Guide YVL 1.5. The guide gives also requirements for the content and submitting schedule of the reports.

In accordance with the Guide utilities are required to submit to STUK reports on a regular basis (daily, quarterly and annual reports; outage reports; environmental radiation reports; reports on individual doses; reports on the utilisation of operational experience) and on event basis (event reports). The types of event reports are:


  • Special reports (events relevant to the nuclear safety, to the safety of the plant personnel or to radiation safety in plant's environment)
  • Reactor scram reports
  • Operational transient reports


Additionally, the utilities are also required to inform immediately STUK about events classified as level 1 or higher on the INES Scale. The proposed INES level shall also be submitted to STUK. A corresponding notification shall be issued also if an event is classified as level 0 but it attracts or might attract public interest. Requirements are set in Guide YVL 1.5.

Incident reporting from STUK to the public

Prompt information

Guidance for the public information are given in an internal guide. Examples of the events to be promptly reported to the public are:


  • The INES level of an incident is 1 or higher or an incident has attracted or might attract public interest.
  • A reactor trip.
  • Exceptional radioactive doses to the personnel or a significant incident from the viewpoint of the industrial safety (e.g. a serious industrial accident) or neglecting the occupational safety aspects.
  • Fire or ignition in consequence of which a fire brigade is called onsite.
  • Exceptional radioactive leakages and exceptional incidents when handling radioactive substances onsite or during transport.


Tools available for an immediate information are press releases and press conferences, TextTV (co-operation with the Finnish Broadcasting Company) and STUK's www-pages.

Later information

STUK issues quarterly reports on the operation of Finnish nuclear power plants. The reports are published within three months after the end of a reporting period. The reports include descriptions of events of which a special report have been submitted to STUK or which have been classified as level 1 or higher on the INES Scale. Single failures are reported if they affect significantly the electricity production of a plant unit or if a common cause failure is in question. Power diagrams describing electricity generation at each plant unit and the causes of power diagrams are also given. The quarterly reports are distributed to Finnish media, state authorities, utilities, research institutes, some libraries, and municipalities near nuclear power plants. The reports are published in whole also on STUK's www-pages.

The quarterly reports are translated into English. Translations are distributed e.g. to regulatory bodies in other countries.

Incident reporting from STUK to international organisations

Finland participates in two international incident reporting systems: IAEA/NEA's INES system (International Nuclear Event Scale) and IAEA/NEA's IRS system (Incident Reporting System). For the both systems, the national co-ordinator has been nominated from STUK.

STUK reports to the IAEA for the INES Information System about the event occurred at Finnish nuclear power plants as agreed internationally, i.e. an Event Rating Form is sent to the IAEA within twenty-four hours if an event has been classified as level 2 or above on the INES Scale or an event has attracted public interest outside Finland.

Events at Finnish nuclear power plants are reported to the IRS system in accordance with the IAEA/NEA's Reporting Guidelines. Events to be reported are decided by STUK. Reports are compiled by STUK and a draft report is circulated for comments to the utility concerned.

Please clarify whether the plants use symptom based accident management procedures? How are these procedures linked to SPDS, to emergency notification criteria and to manuals for severe accident management?

Loviisa plant

The symptom based emergency operating procedure (EOP) was developed in the middle of 1980's. SPDS was created approximately at the same time. The goal was to make both systems as identical as possible. For example, most of the limit values are same in both systems. SPDS, the critical safety function monitoring system as it is called, contains logics for line-up of safety systems and recognition of accidents like the spurious opening of a pressurizer safety valve, steam generator tube rupture and leakage in the containment. Severity categories for the safety functions are an essential feature of SPDS.

The main safety functions are the same in both systems, but the EOP contains more items to be monitored like checking alerts and reading some local measurements. Many of the emergency notification criteria are given in the EOP concerning of general emergency. Examples of these criteria are:


  • core exit temperature > 450°C
  • dose rate in containment > 50 mSv/h
  • activity in the containment, noble gases > 107 kBq/m3
  • activity concentration of the effluent air in the stack > 0.1 GBq/m3 (Kr -87 eqv)
  • dose rate at the plant site > 0.4 µSv/h
  • containment overpressure > 0.7 bar


Olkiluoto plant

Symptom based emergency operating procedures (EOPs) were written for the plant in the late 1980´s. The EOPs consist of a general procedure for identification of symptoms and supervision of progression of the safety critical parameters, nine symptom specific procedures, a special procedure for mitigation of severe reactor accidents and another special procedure for re-establishing power supply in the case of a station blackout.

The EOPs were already available when the design of the SPDS system was started, and they formed a natural basis for the set-up of the display system. In principle, there is one specific display for each EOP, so that all the information needed to cope with a certain symptom can be seen on the screen simultaneously. The SPDS system also alarms the operators each time a symptom is “activated”, shows the status of the reactor protection system (actuated conditions) etc.

If the reactor cannot be made subcritical or if a low water level cannot be restored within a certain time, the corresponding EOPs will guide the operators to apply the ultimate procedure for mitigating the consequences of severe reactor accidents. This ultimate procedure also provides guidance to the emergency preparedness activities.

The Report does not give any particular examples on how the industry operating experience of similar reactor types is taken into account. It is desirable that this matter be disclosed?

Loviisa plant

The Loviisa plant is a VVER-440-type reactor and even before the start of operation in 1977 there has been continuous cooperation with other VVER-operators. There has been exchange of information between two plants (for example Loviisa-Paks) and also meetings between plant managers concerning common problems. From 1989 very important tool in cooperation has been WANO (World Association of Nuclear Operators) and its Moscow Centre. Loviisa has been an active member in Moscow Centre and whole WANO.

Also the annual retraining to plant personnel on foreign nuclear incidents include operating experience of similar reactor types.

Olkiluoto plant

The operating experience of similar reactor types is followed by several means. The main sources of information are ERFATOM, KSU and Forsmark which are explained in more detail below. Information is also coming directly from several sources (WANO, IAEA and OECD/NEA (IRS), Loviisa power plant (e.g. operating experience meetings and reports), vendors (ABB Atom, ABB Stal), component manufacturers, WANO Network, BWROG (BWR Owners Group). Screening of events relies to a large extent on KSU and ERFATOM.

The group for operating experience feedback consists of 6 members on site. This group gives recommendations to the organisation which decides the possible corrective actions.



  • founded by the Swedish utilities and TVO as a consequence of the so-called Barsebäck incident (July 1992)
  • activities started on January 1st, 1994 in the premises of ABB Atom (Västerås, Sweden)
  • information sources more “domestic” than those of KSU
  • issues weekly reports, monthly reports, annual reports and topical reports
  • gives recommendations
  • event classification system


  • Swedish nuclear training and safety centre
  • interface between WANO/INPO (membership of INPO has been cancelled in 1999) and the Swedish NPPs
  • regular screening of international operating experience information from the Swedish (and TVO's) point of view
  • screening results are reported regularly in the monthly reports of KSU
  • ERFNOVA database


  • Forsmark units 1 and 2 in Sweden can be called as “sister units” of Olkiluoto 1 and 2.
  • reports from sister units (e.g. licensee event reports)
  • minutes of the meetings of safety committees.


1. Teollisuuden Voima Oy – organisation2. Fortum's main subsidiaries

3. Fortum's Power and Heat Division

Kalervo Nurmimäki (starting on 1 January 2000 Tapio Kuula)
Rauno Kallonen
Teemu Järvenpää
Tapio Kuula
Eero Auranne
Control Centre Trading
Sales to large- scale customers
Sales strategy
Fortum Energy Partners
Neste Lämpö Oy
Koillis-Pohjan Sähkö
Uudenmaan Energia
Minority holdings
UK + Ireland
Baltic countries
Technology Centre
Business planning
Fortum Advanced Energy Systems


4. Organisation of the Power Business Unit

5. Loviisa Power Plant organisation

6. Collective occupational radiation doses to NPP workers

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