BfS SAFETY CODES AND GUIDES - TRANSLATIONS. Edition 1/01. Contents. Bundesamt für Strahlenschutz Salzgitter

Transcription

1 BfS SAFETY CODES AND GUIDES - TRANSLATIONS Edition 1/01 Contents RSK Guidelines for Pressurized Water Reactors, 3rd Edition of 14 October 1981, amended 1982, 1984 and 1996 RSK - Leitlinien für Druckwasserreaktoren 3. Ausgabe vom 14. Oktober 1981, ergänzt 1982, 1984 und 1996 Bundesamt für Strahlenschutz Salzgitter

2 he German original of this translation was published in Bundesanzeiger (Federal Bulletin, BAnz.) 1982, No. 69 a, the amendments were published in BAnz. 1983, No. 106, BAnz. 1984, No. 104, BAnz 1996, No. 158 a and BAnz 1996, No. 214 In case of discrepancies the German text shall prevail.

3 RSK Guidelines for Pressurized Water Reactors 3rd Edition of 14 October 1981, amended on 15 December 1982, on 21 March 1984, on 20 March 1996 and on 29 October 1996 Preface 1. Definition 2. Site 2.1 Site Evaluation 2.2 Assumptions for the Determination of the Post- Incident Radiation Exposure of the Environment 3. Reactor Core 3.1 Reactor Design Nuclear and Thermohydraulic Design Shutdown Systems Operational Transients Requiring the Reactor Scram Pressurizer Relief Valves/Pre-Closing 3.2 Inherent Safety 3.3 Reactor Pressure Vessel Internals 4. Vessels and Pipe Systems 4.1 Pressure-Retaining Boundary of the Reactor Coolant Scope of Application Design, Layout and Choice of Material Fabrication Principles Destructive In-Process Tests Nondestructive In-Process Examinations Operation Principles Destructive In-Process Tests (Surveillance Test Specimens) Recurrent Pressure Test Recurrent Nondestructive Examinations of the Reactor Pressure Vessel Examination Technique Required Examination Scope Permissible Discontinuities (being prepared) Examination Intervals Recurrent Nondestructive Examinations of Other Components and Systems of the Pressure-Retaining Boundary 4.2 External Systems Scope of Application Requirements for Construction, Design, Materials, Manufacture and Testability Manufacture Operation 5. Containment Vessel 5.1 Design Bases 5.2 Structural Layout 5.3 Design Conditions and Material Requirements 5.4 Manufacture 5.5 Leak Tests, Leak Rate Tests 5.6 Penetrations 6. Electrical Equipment of the Process System 6.1 Process Monitoring 6.2 Process Control Equipment 7. Electrical Equipment of the Safety System and the Other Safety-related Systems 7.1 Scope 7.2 General Requirements Design Inspections Documents to be Presented 7.3 Instrumentation and Control Control Systems Important for Safety Scope General Requirements Requirement Specification Registration of Accidents and Disturbances Redundancy and Independence Qualification Qualification of the System Qualification of the Equipment Robustness Man-Machine Interface Inspection, Maintenance 7.4 Electrical Components of the Safety System and of the Other Safety-related Systems 7.5 Electrical Energy Supply of the Safety System and of the Other Safety-related Systems 7.6 Safety Instrumentation and Control Software Requirements to be met by the Provision and the Examination of Software Software for the Safety Instrumentation and Control of Categories 1 to Safety Instrumentation and Control Software of Category Principles Constructive Quality Assurance Analytical Quality Assurance Organization and Administration The Use of Standard Software Software for the Safety Instrumentation and Control of Category Principles Constructive Quality Assurance Analytical Quality Assurance Organization and Administration The Use of Standard Software Software for the Safety Instrumentation and Control of Category Principle Constructive Quality Assurance Organization and Administration The Use of Standard Software Requirements for Operation and Ensurance 8. Control Room 9. Ventilation System 10. Radiological Protection Monitoring 10.1 In-Plant Radiological Protection Monitoring Monitoring of the Local Dose Rate Monitoring of Compartment Air Radioactivity Loop Monitoring Detection of Contamination 10.2 Activity Monitoring Liquid Waste Activity Monitoring Exhaust Air Activity Monitoring 11. Fire Protection 12. Escape Routes and Alarms 13. Arrangement of Work Cycle, Work Area and Work Environment 14. Arrangements for Work to be Carried out at the Plant 15. Handling and Storage of Nuclear Fuels and other Radioactive Substances 16. Decommissioning and Disposal 17. Provisions against Component Failure 17.1 Provisions against Turbine Failure 17.2 Provisions against Pump Flywheel Failure 17.3 Provisions against Backflow Preventer Failure 18. Natural Events 18.1 Earthquake RSK Guidelines for Pressurized Water Reactors 1

4 19. Man-Induced Events 19.1 Aircraft Crash 19.2 Chemical Explosions 19.3 Poisonous and Explosive Gases 19.4 Events Caused by Third Parties 20. Failure of the Scram System during Operational Transients 21. Leaks and breaks to be postulated 21.1 Leak areas to be postulated in the reactor coolant pipe, including austenitic connecting pipes (grade steel) of DN > 200 mm, and in the reactor pressure vessel 21.2 Leaks and breaks to be postulated in the main steam and feedwater pipes 22. Systems for Post-Incident Heat Removal 22.1 Emergency Core Cooling and Residual Heat Removal System Requirements Design Assumptions for Emergency Core Cooling Calculations 22.2 Emergency System 23. Remote Operated Degasifying of the Primary System 24. Measures for the Limitation of the Hydrogen Concentration 25. Incident Instrumentation 25.1 General Requirements 25.2 Incident Course Instrumentation 25.3 Incident Sequence Instrumentation Design Functional Tests Preface The Reactor Safety Commission prepared the following Guidelines as a compilation of the safety-related requirements which, in the Commission's opinion, have to be complied with in the design, construction and operation of a nuclear power plant with a pressurized water reactor. This third edition of the RSK Guidelines for Pressurized Water Reactors is a reviewed and enlarged version of the second edition of January 24, As of October 1981, the Reactor Safety Commission will use the present Guidelines as a basis in its deliberations concerning site and safety concept of pressurized water reactors prior to the issuance of a construction permit, and it will measure the construction phases of the corresponding nuclear power plants against the background of these Guidelines. They are not readily intended for an adaptation of nuclear power plants which do already exist or are under construction or in operation. The scope of application of the Guidelines to these plant will have to be examined on a case-by-case basis. The major purpose of the Guidelines is a facilitation of the discussion process within the Reactor Safety Commission and an early submission of references to safety-related requirements which the Reactor Safety Commission considers necessary. Should manufacturer and licensee comply with the Guidelines the Reactor Safety Commission will proceed with the short-term issuance of comments on individual projects. If the applicant is unable or unwilling to comply with a certain guideline he will have to demonstrate that other measures will assure safety in at least an equivalent way. The main idea behind this provision is to allow the necessary latitude in the further development of safety technology. The Guidelines have been issued as a loose-leaf collection which will require to be updated at adequate intervals in order to make allowance for both the then prevailing state of the art on which to base the concept of a nuclear power plant and the further development of engineering codes and guides. Some chapters of the Guidelines are rather comprehensive as corresponding codes are still missing. Where appropriate codes or guides have already been issued, the present text contains a corresponding reference. Cologne, October 14,

5 1. Definition (1) Safety System The safety System is the entirety of the features of a reactor plant which are to protect the plant against inadmissible stresses and keep within given limits the impact which an incident may have upon operating personnel, plant and environment. 2. Site 2.1 Site Evaluation The following BMI Guideline is available: "Compilation of the Information Required for Review Purposes under Licensing and Supervisory Procedures for Nuclear Power Plants" A. Site, Issue of October 20, The following shall apply in addition: (1) The expert analysis and opinion report on the radiation exposure of the public in the environment of the site shall discuss the following items: 1. Radiation exposure as a result of the applied-for discharge of radioactive substances from the projected nuclear power plant during specified normal operation. 2. Calculation of the radiation exposure in the environment of the applied-for site as a result of the discharge of radioactive substances from other nuclear facilities and from the handling of radioactive substances. The calculations shall be done on the basis of the principles of calculation, as amended, which have been issued by the Federal Minister of the Interior for the implementation of Sec. 45 of the Radiological Protection Ordinance. Nevertheless, a summary of the principles of the calculation of the radiation exposure of the public in the environment of the site shall be submitted. In particular, a complete list of the exposure pathways discussed and the computer codes used, including input data, shall be provided so as to have available an unequivocal documentation of contents and scope of the calculations. 2.2 Assumptions for the Determination of the Post- Incident Radiation Exposure of the Environment (1) Suitable design measures shall be taken to keep the radiological consequences of incidents as low as possible within the permissible limits under Sec. 28, para. (3), first sentence of the Radiological Protection Ordinance. Allowance shall be made in this context for the fact that the occurrence probability of these incidents must be sufficiently low. Releases during abnormal operation (e.g. a short-term loss of station service power) shall be accounted for within the operational discharges under Sec. 45 of the Radiological Protection Ordinance. (2) Analysis shall be performed to determine the relevant incidents which may lead to a release of radioactive substances to the environment of the nuclear power plant. (3) With regard to the incidents determined in accordance with para. (2) above, the radiation exposure of the environment shall be calculated using realistic and sufficiently confirmed assumptions concerning the release of fission products into the containing systems, the factors of deposition on the internals as well as the time history of leak and/or blowdown rates of the containing systems. (4) In particular, the calculation of the radiation exposure after the postulated break of a reactor coolant pipe shall be based on the following hypothetical assumptions: 1. Core inventory The calculation of the activity inventory during an incident shall be based on an equilibrium core at the end of an operating cycle. Thereby for the third of the core with the highest burn-up, a full power operation of 2.5 years shall be assumed. 2. Scope of the release of radioactive substances from the fuel rods It shall be assumed that 10% of all fuel rods fail, as far as no lower value is proved by a damage extent analysis. In relation to the inventory of a fuel rod, the following percentages shall be assumed as having been released spontaneously: 10 % of the noble gases 3 % of the halogens 2 % of the volatile solids (Cs, Te) 0.1 % of the other solids. Airborne materials in the containment vessel The following percentage of spontaneously released airborne radioactive materials existing in the containment vessel shall be assumed with regard to deposition, washout and similar effects during transportation to the containment vessel and inside the containment vessel 100 % of the noble gases 25 % of the halogens 2.5 % of the solids The airborne radioactive materials shall be assumed as being equally distributed in the containment atmosphere, while the washed-out and deposited percentage of the fission products and activated products shall be assumed as being in the sump water. 3. Chemical form of airborne halogens It shall be assumed that after completion of the separation processes specified in subpara. 3 above the airborne halogens then still existing in the containment vessel will be made up of 85% elementary halogens, 10% organically bound halogens and 5% aerosols. 4. Containment vessel leak rate The calculations of the time slope of the leak rate may be based on the pressure-time function resulting from the incident analyses. The leak rate course shall be calculated using the determinations of the KTA Safety Standard 3405: "Integral Leakage Rate Testing of the Containment Vessel with the Absolute- Pressure Method", Issue 2/79. If the timely variation of the pressure should not be considered for further calculation simplification, the design leak rate shall be applied constant in a conservative way for the duration of 24 hours. After this time period no further leak rate need be postulated. 5. Percentage of organic iodine in the filter supply air During its transportation to the filters, part of the elementary iodine is converted into organically bound iodine. Thus the total iodine in the filter supply air shall be assumed to contain 50% organic iodine. 6. Separation efficiencies of incident filters The following separation efficiencies shall be postulated with regard to the effectiveness of the filters intended for the control of the incident: 0 % for noble gases 99 % for organically bound halogens % for elementary halogens 99.9 % for all aerosols The requirements for the filter devices for incident control are laid down in Sec. 9 Ventilation Systems. 3

6 3. Reactor Core 3.1 Reactor Design The following KTA Safety Standard is available: KTA : Part 1: Design of Reactor Cores for Pressurized Water and Boiling Water Reactors Principles of Thermohydraulic Design, Issue 2/ Nuclear and Thermohydraulic Design (1) The heat flux densities upon which the thermal design of the reactor core has been based shall definitely not be exceeded during specified normal operation. A sufficient supervision of compliance with these limits shall be possible by way of measurements. (2) The reactor shall be designed so as to avoid film boiling during specified normal operation with an adequate safety margin. The relations used for the determination of the DNB ratio shall have been sufficiently confirmed by measurements. (3) Fluctuations in operating variables during specified normal operation (power, flow rate, temperature, pressure) shall not cause any noncompliance with the limits specified as permissible limits for the fuel assemblies and for those components which are of importance for core integrity. (4) In the design of the fuel assemblies and their calculation, stringent requirements for calculation bases and calculation methods shall be met considering all significant effects such as the influence of radiation on the properties of materials, chemical processes, static and dynamic, mechanical and thermal stresses, dynamic (structural) behavior of the fuel and cladding tube materials as well as calculatory uncertainties Shutdown Systems (1) The reactor shall have two independent and diverse shutdown systems. Either shutdown system shall be capable of bringing the reactor into the subcritical state from any stationary operating state without exceeding the permissible operational limits of the reactor core. (2) At least one of the two shutdown systems (referred to as scram system) shall be capable of bringing the reactor into the subcritical state so quickly - even in case of possible reactivity perturbations - that the given permissible limits will not be exceeded. At least one of the two shutdown systems, when used alone, shall be capable of keeping the reactor core in a long-term subcritical state. (3) At least one shutdown system shall be capable of preventing the specified limits from being exceeded during anticipated transients and, together with other safeguards, of preventing damage to the reactor core and the pressure-retaining boundary during incidents. These conditions shall also be complied with if, as a result of a malfunction of any kind whatsoever, there is a failure of the signal providing the first activation. (4) A scram system (cf. para. (2) above) shall be provided that will be released by activations which, as a matter of principle, are derived from different process variables and are redundant among themselves. If, in certain incidents, activations can only originate from a single process variable care shall be taken to assure a higher safety level in the design of the detection and transmission of the corresponding signal as compared with the detection and transmission of the signals of the other variables. (5) The drives of the control assemblies, including all relevant auxiliary systems, shall have common components only insofar as it has been made sure that a single failure will not affect the reliable shutdown of the reactor. (6) The shutdown systems shall be designed in such a way that the shutdown of the reactor is assured if several events occur which are not independent of each other (e.g. fires, ingress of water). (7) Shutdown systems shall be capable of such a fast compensation of the reactivity gain at maximum cooldown rate (also in case of incidents such as a main steam pipe rupture or an ingress of cold water) that the reactor core can be transferred into and kept in the subcritical state. A brief recriticality will be acceptable in this context provided certain specified limits are not exceeded. (8) If one of the two shutdown systems uses a boron injection system a monitor shall be provided to supervise the boron concentration in the supply tanks which are important to safety. (9) The shutdown system using the control assemblies shall be designed in such a way that the calculated shutdown reactivity will not fall below 1% during the entire core life at zero load in the hot operating state and with the most effective control assembly withdrawn. (10) Measures shall be taken to prevent an uncontrolled withdrawal of control assemblies. It shall be demonstrated that such an incident will nevertheless be controlled if it does occur. (11) To cope with the ejection of a control assembly which may occur as a result of a guide tube or drive fracture, a second independent safeguard such as an intercepting device shall be provided apart from the safe design and thorough fabrication control of the guide tube unless the energy release resulting from the ejection of the control assembly with the greatest reactivity value will definitely not cause any damage to the reactor core and the reactor system. In this case, the applicant shall submit the corresponding evidence. The fracture of a control assembly guide tube or a control assembly drive shall not cause any consequential damage to adjacent control assembly guide tubes or control assembly drives which would impair the functional reliability of other control assemblies. If such consequential damage cannot be precluded it shall be demonstrated that even such an incident will not cause any damage to the reactor core. (12) If the scram system and the control system have certain components in common (e.g. control assemblies) it shall be assured that none of the functions of the control system (including those which may originate from failures in the control system) will prevent or affect the specified normal function of the scram system. (13) At regular intervals, all control assemblies shall be tested for easy and smooth motion and the rod insertion times shall be recorded. (14) If a control assembly can no longer be moved the reactor shall be made subcritical and examined with a view to further operability Operational Transients Requiring the Reactor Scram (1) An analysis shall be performed on the development of transients anticipated during the life of the reactor (operational transients) and generating such changes in operating variables that will cause a reactor scram. (2) It shall be demonstrated that, with regard to the transients discussed, 1. the heat flux will be sufficiently far from the critical heat flux; 2. the pressure within the pressure-retaining boundary will remain in principle below the response pressure 4

7 of the safety valves; and 3. the energy release inside the fuel rods is so low that no melting will occur. The analyses shall be based on such initial conditions as are typical for the type of investigations concerned (e.g. overload or failure of the first scram activation) Pressurizer Relief Valves/Pre-Closing (1) The plant shall be designed procedurally in that way that only during infrequent transients with high increase of pressure an actuation of the pressurizer valves is to be expected. (2) The relief valves shall be equipped with a pre-closing, which closes automatically when the valve is stuck open faulty. (3) For the initiation of the pre-closing (detection of the faulty stuck open relief valve) direct information about the relief valve position shall be used. 3.2 Inherent Safety (1) Besides the negative reactivity coefficient of the fuel temperature, the reactivity coefficient of the coolant temperature during nominal operation shall also be negative. If the reactivity coefficient of the coolant temperature is not negative for a certain time during the start-up phase, it shall be demonstrated with a view to incidents (loss of coolant) that the additional power contribution is harmless. (2) With regard to nominal operation it shall be demonstrated that any local void formation which may occur in incidents involving depressurization will have a negative effect upon reactivity and local power density. (3) If a temporary positive effect cannot be precluded for states below nominal temperature (cycle start), it shall be demonstrated with a view to incidents that the additional power contribution is harmless. 3.3 Reactor Pressure Vessel Internals (1) The reactor pressure vessel internals shall cope with the stresses generated during specified normal operation. The reactor pressure vessel internals shall be designed in such a way that, with regard to the incidents referred to in Sec. 21.1, the subcriticality of the reactor will be assured by the scram system (drop of a sufficient number of control assemblies) and the long-term shutdown condition as well as the coolability of the reactor core will be assured by the systems intended for this purpose. (2) Adequate measures shall be taken for the detection of loose or detached particles within the pressure-retaining boundary. In addition, adequate measures should be taken for the exact localization of such loose or detached particles within the pressure-retaining boundary. (3) The vibrational behavior of the reactor pressure vessel internals shall already be examined by suitable measurements during the commissioning phase of the plant. Repetitions of such measurements shall also be possible during operation of the plant. 4. Vessels and Pipe Systems 4.1 Pressure-Retaining Boundary of the Reactor Coolant The following KTA Safety Standards are available: "Components of the Reactor Coolant Pressure Boundary of Light Water Reactors" KTA Materials, Issue 2/79 KTA Design and Analysis, Issue 10/80 KTA Manufacture, Issue 10/ Scope of Application (1) The pressure-retaining boundary of the reactor coolant includes the reactor pressure vessel, those parts of the steam generators which contain primary coolant, the pressurizer, the reactor coolant pumps and the connecting pipes and nozzles including the first isolating valve. The secondary shell of the steam generators shall be treated as a pressure-retaining boundary with regard to selection of material, design principles, quality assurance, inprocess inspection and recurrent tests and examinations. The same shall apply analogously to pipes of a small nominal bore such as instrument or control pipes unless safety-related consequences can be excluded if such pipes fail Design, Layout and Choice of Material (1) The following requirements will provide the pressureretaining boundary with a basic safety that will preclude any disastrous failure of a plant component as a result of manufacturing defects. The basic safety of a component depends on the following features: high-grade material properties, in particular ductility; conservative limitation of stresses; prevention of stress peaks by way of optimized design and construction; assurance of the application of optimized manufacturing and testing technologies; knowledge and assessment of fault conditions, if any; consideration of the operating medium. The following detailed requirements will have to be complied with: (2) The pressure-retaining boundary shall be designed in such a way that it can be subjected sufficiently often to the maximum stresses occurring during specified normal operation and incident conditions. (3) The structural layout of all components of the pressure-retaining boundary shall be such as will enable examinations and tests during manufacture and at the site of installation as well as recurrent examinations and tests to a sufficient extent. This applies in particular to the welds. (4) The materials of the pressure-retaining boundary and their cladding shall be such as will enable sufficient nondestructive examination of all components in accordance with the present Guidelines. (5) All those areas of the pressure-retaining boundary where recurrent tests and examination are impossible or restricted shall be so small that postulated defects of the size of these areas cannot result in a component failure whose consequences would be beyond control. (6) The choice of material and adequate structural design, welding and heat treatment shall assure a sufficiently ductile state of the material throughout the life of the plant at all points of the pressure-retaining boundary under all plant conditions which may occur during normal operation and incidents. This shall be assured, among other things, by limiting the maximum neutron fluence in the core area close to the wall of the reactor pressure vessel to 1 x cm -2 (energy 1 MeV). As far as those areas of the pressure vessel wall are concerned which are exposed to neutron radiation, the basic material and the weld metal shall comply with all requirements for a chemical composition that will reduce radiation embrittlement. (7) As a matter of principles, all details of the requirements for design, layout, material and manufacturing processes shall be agreed upon between reactor vendor, component manufacturer and authorized 5

8 inspection agency. Design, choice of material and processing steps shall be reviewed by the authorized inspection agency under consideration of the state of the art. (8) For all components of the pressure-retaining boundary, only such materials shall be used as have been certified with regard to manufacturer, product form and manufacturing process. This certification shall be completed at the latest at the time of commissioning or shall be covered by a single certification procedure for the component concerned. 1. For ferritic materials (other than bolt steels) it shall comprise, among other things, a review of the segregation behavior and investigations of weldability and welding safety including, among other things, welding simulation tests, round welding-in tests in case of thick-walled product forms as well as tangential sectioning tests. The purpose of these investigations of simulated welding specimens is the detection of changes in the ductility behavior of structures of the heat-affected zones following thermo-mechanical stresses in the temperature range of stress-relief annealing. The minimum value of the energy absorbed in an impact test shall be 68 J, as determined on specimens simulated under close-tofield conditions, and it shall not be lower at any temperature of the operating states. 2. The component ductility behavior shall be analyzed by means of large-scale specimens and componentlike tests in consideration of damaged and defective material areas. 3. In addition, the long-term influences of vibrational stresses and the corrosion effects resulting from static and cyclic stresses shall be investigated including the conditions under which stress corrosion cracking occurs. 4. All investigations shall be performed on a sufficiently broad parameter basis. 5. With regard to materials where other investigations, comprehensive fabrication experience and longstanding successful operation are evidence that they are suited for components of the pressure-retaining boundary even in the light of the aforementioned aspects, the condition of certification shall be deemed to have been complied with if the authorized inspection agency issued an explicit certificate to this effect including justification 1. (9) Ductility requirements 1. The ferritic materials shall have such properties that, for base materials, weld metals and heat-affected zone, the reference NDT temperature is at least 33 K below both the lowest operating stress temperature and the pressure test temperature. In this context, the reference NDT temperature is defined by the following measures: Establishment of a temperature T NDT which is equal to or higher than the nil ductility transition temperature, determined with the aid of drop weight tests. At a temperature that is not higher than T NDT + 33 K, each specimen from the notched bar impact test (ISO-V transverse specimens) shall have at least 0.9 mm lateral expansion and no less than 68 J of absorbed energy. If these requirements are complied with, temperature T NDT is equal to temperature RT NDT. The absorbed energy in the upper plateau (ISO-V transverse specimens) at the point where the specimen was taken shall attain at least 100 J in a depth of one-fourth of the Q and T section thickness (individual value). In case the above requirements are not complied with additional notched bar impact tests (ISO-V transverse specimens) shall be performed in sets 1 Until further notice these certificates shall be agreed upon with the Reactor Safety Commission. of three specimens each, to determine the temperature T AV at which the above requirements will be met. In this case, reference temperature is RT NDT = T AV - 33 K. Thus, the reference temperature RT NDT is the higher of T NDT and T AV - 33 K. If the notched bar impact test was not carried out at T NDT + 33 K or if it does not provide the minima of 68 J and 0.9 mm lateral expansion at T NDT + 33 K, the temperature at which the minima of 68 J of absorbed energy and 0.9 mm of lateral expansion are attained shall be determined on the basis of the complete energy absorbed temperature curve (a v -T curve) which will be established from the lower values of all specimens. 2. As far as the notched bar impact tests are concerned, a test may be repeated at the original temperature provided the following conditions are complied with: The mean test result shall not be lower than the specified individual value. No more than one specimen may be below the specified individual value. The results of the specimen which has not attained the specified individual value shall not be below the specified value by more than 13.6 J or 0.13 mm. 3. A retest shall comprise two additional specimens for the one that failed; the specimens shall be taken at a point that is as close as possible to the point where the original specimen has been taken. The repeated test shall only be considered as successful if both specimens attain the specified individual value. 4. If the demonstration of the NDT temperature is not possible because of the component geometry (thinwalled product form) it shall be demonstrated that the energy absorbed is at least 68 J and the lateral expansion at least 0.9 mm, provided the dimensions of the component enable the taking of a notched bar specimen. Although, as a rule, these values shall be attained with ISO-V transverse specimens, specimens similar to ISO-V specimens may be used for assessment purposes in case of small wall thicknesses. 5. For non-ferritic materials (martensitic, austenitic), the ductility requirements shall be agreed upon with the authorized inspection agency within the frame of the manufacturing and test schedules. (10) To complement the material certification, the ductility behavior after long-term stresses (temperature, load change, neutron irradiation and corrosion) shall be determined on smallscale specimens for the evaluation of long-term behavior. (11) As a rule, seamless tubes shall be used for the straight legs of the pipes of the pressure-retaining boundary. Seamless bends should be used wherever possible Fabrication Principles (1) Fabrication and its supervision shall be made part of a formal licensing process right from the beginning. Prior to the beginning of fabrication, the component manufacturer shall prepare manufacturing and test schedules. These shall be coordinated with the reactor vendor and the authorized inspection agency. Component manufacturer, reactor vendor and authorized inspection agency shall see to it that the flow of manufacture is supervised and documented in its entirety. Individual provisions concerning destructive tests and nondestructive examinations follow: (2) Any and each elimination or repair of a defect shall 6

9 require the prior consent of the authorized inspection agency or its nominee. In this context, it shall be considered in how far the intended repair will involve safety-related advantages or disadvantages as compared with the unrepaired state. Work processes and characteristic fault data shall be documented. (3) Suitable certification and/or qualification tests shall be carried out with regard to weld metals, filler metals and welding consumables. (4) Weld claddings shall be capable of nondestructive examinations and shall be such that the basic material can be subjected to a perfect ultrasonic examination in and out. (5) As far as cast steel casings are concerned, all casting details which are important in terms of quality and faultfree condition shall be included in the fabrication schedule. Fabrication weldings of ferritic steels shall be normalized or tempered or stress-relieved, depending on the material concerned, and shall be documented Destructive In-Process Tests (1) It shall be demonstrated by suitable tests that there are sufficient ductility and strength properties over the entire wall thickness. (2) If with regard to components having a great wall thickness the minimum values of yield strength and tensile strength, as specified in the material standards, are not attained in the interior of the wall, such shortcomings shall only be permissible if they do not amount to e.g. more than 10 % in the cylindrical area and more than 20 % in tube sheets and flange rings and if it is demonstrated that the energy absorbed in a notched bar impact test will reach at least 68 J in the middle of the wall at a test temperature of 80 C. (3) Compliance with these requirements shall be demonstrated not only by the investigations within the frame of the expert analysis, but also to a representative extent in the course of the in-process tests and examinations if this is possible, with regard to semifinished products, by taking test cores, e.g. in the area of nozzle connections and the like. (4) Within the frame of the procedure and production tests, extended metallographic tests (transverse micro and multiplestage tangential sectioning over the entire wall thickness) shall be carried out to a representative extent. These tests shall also be performed for repair weldings. Moreover, welding simulation tests shall be performed to a representative extent with regard to the melts employed. These extended tests may be waived within certain analysis limits if it is demonstrated for these analyses, within the frame of the expert analysis of the materials under Subsec , para. (8), that no relaxation embrittlement and no relaxation cracks will occur. However, if contrary to the results of these investigations cracks or indications of embrittlement or cracking occur during fabrication, tangential sectioning tests and welding simulation tests will again be necessary. As far as the material of the shell of the reactor pressure vessel is concerned (with the exception of nozzles and welded attachments), simulated welding specimens shall be produced from each heat during manufacture and shall be tested in accordance with Subsec , para. (8) Nondestructive In-Process Examinations (1) All product forms intended for the pressure-retaining boundary shall be subjected to nondestructive examinations involving a sufficient detectability of faults. The examinations shall include the ultrasonic examination of each volume element and a complete surface crack examination. With regard to greater pipe lengths having nominal bores of less than 500 mm, the surface crack examination of the inner surfaces shall be performed to the extent these surface are accessible from the pipe ends. (2) The examination methods and sound scanning directions shall be selected in such a way that all safetyrelated discontinuities will be localized irrespective of their orientation. In this context, discontinuities whose orientation is vertical to the main stress directions (operational stresses) shall be considered by selecting particularly suited examination techniques and sound scanning directions. For example, the examination of forgings, sheet metals, castings, pipes having an outer diameter of 100 mm or more and tubes for control rod penetrations shall be done by both straight beam and angle beam scanning. In case of angle beam scanning, at least two different examination directions shall be used which should differ by about 90 if the examination is carried out from two directions only. When forgings having a wall thickness of more than 100 mm are subjected to ultrasonic examinations, the tandem technique or a similar examination technique shall be used in addition. Pipes having an outer diameter of less than 100 mm shall be examined at least for longitudinal discontinuities. The surface crack examination shall cover all directions. (3) As regards the thick-walled product forms of the wall of the reactor pressure vessel, the pressurizer, the steam generators (including the secondary shells) and the inserted nozzles, the sound transmission shall be considered in addition. This means that wherever a back reflection or shape reflection is anticipated as a result of the prevailing geometry, such reflection will actually occur. Where this echo disappears or is materially weakened the corresponding locations shall be examined with refined methods to discover the cause of this impaired sound transmission. (4) Cast steel valve casings shall be subjected to a complete radiographic examination instead of the ultrasonic examination described in paras. (1) and (2) above. In case of ferritic steel castings, ultrasonic examinations shall be performed on fabrication weldings, welded attachments and such locations where radiography will not supply the necessary exposure quality or where these examinations are required because of structural or casting particularities. In case of austenitic steel castings, restrictions of radiographic examinations shall be avoided by appropriate structural measures. (5) Prior to welding, a surface crack examination of the weld side walls shall be carried out. In case of ferromagnetic materials, the examination shall be carried out with magnetic particles. (6) After their last heat treatment, all welds shall be subjected to an ultrasonic examination and one of the techniques of the surface crack examination providing a sufficient detectability of discontinuities. As to welds on components having a nominal bore of less than 500 mm, the ultrasonic examination may be substituted by radiography if the sensitivity of the ultrasonic examination is impaired to such an extent by the properties of the material that even the application of the techniques of the pulse reflection method which are available in accordance with the state of the art - such as single search unit technology, focusing search units (with curved lenses), transmitter-receiver technology, wide band search units, application of longitudinal waves - will not supply any meaningful result. (7) In case of components having an outer diameter of 100 mm and more, the welds shall be examined again after the pressure test. Details shall be agreed upon with the authorized inspection agency. For circumferential welds on pipes, this examination may be waived if no significant stresses have occurred vertically to the weld. (8) Should it become obvious from the fabrication sequence that the examinations after the last heat treatment can only be carried out with a lower detectability 7

10 of discontinuities than before, an additional examination shall be carried out at an earlier time at which the prevailing conditions will still assure an optimal detectability of discontinuities. For butt welds having a wall thickness of 40 mm or less, an additional radiographic examination shall be performed at a suitable time within the fabrication sequence. In case of circumferential welds having an outside diameter of less than 100 mm and a ratio of less than 0.1 between nominal wall thickness and outside diameter, the ultrasonic examination may be waived upon agreement with the authorized inspection agency provided the radiographic examination following the last heat treatment has supplied an optimal image quality. (9) The ultrasonic examination of butt, nozzle and attachment welds shall use at least two different sound angles for each volume element in case of wall thicknesses of 40 mm or more. As far as the sound direction is concerned (projection of the main beam onto the tangential plane), all discontinuity orientations vertical to the surface shall be detected. The discontinuity directions which proceed parallel with or at a right angle to the weld or, in case of large-pool weldings, at an angle of 45 to the direction of the weld, shall be taken into account by moving the search unit to and fro whilst leaving the sound direction more or less unaltered. Examinations for discontinuity orientations between the aforementioned directions shall be performed by additional fan-like to-and-fro movements of the search unit. For wall thicknesses of 100 mm and more, the tandem technique shall be applied in addition to the customary single unit technique. (10) If one-single welding cannot be avoided, e.g. in case of pipe assembly welds, detectability as required in Subsec shall be assured without restriction for the ultrasonic examination, among other things by exact adaptation, prevention of disturbing weld offsets, even root formation and sufficiently large coupling surface on either side of the weld. (11) Build-up welds shall be subjected to a complete surface crack examination following their last heat treatment. Austenitic build-up welds shall be examined by straight beam scanning and ferritic build-up welds by additional angle beam scanning after the last heat treatment. (12) Auxiliary welds, i.e. welds which are produced for assembly purposes and removed again, shall be realized in such a way that the penetration into the base material will be limited to the usual weld metal zone. The material properties in the weld metal and in the heat-affected zone shall include a ductility similar to that of joining welds. All points of auxiliary welding shall be documented with an exact statement as to their location and extent even if the auxiliary welds have been removed again. After the last heat treatment they shall be subjected to a hardness test, a surface crack examination and an ultrasonic examination so that discontinuities below the surface will be detected, and they shall be made part of the recurrent inspections unless an etching test has supplied adequate assurance to the effect that the overheated area of the heat-affected zone has been removed completely by dressing. (13) After the last heat treatment the entire surface of the accessible area of the pressure-retaining boundary shall be examined by one of the techniques of surface crack examination. In case of local annealing the examination shall be restricted to the area affected by such annealing. (14) With regard to austenitic steel castings, suitable examinations shall be carried out (e.g. component metallography) to demonstrate that the micro structure is satisfactory. This shall be done component-wise in case of larger components and batch-wise in case of smaller components Operation Principles (1) Sufficient possibilities for inspections and recurrent tests shall be provided for all parts of the pressureretaining boundary. Thus a schedule of recurrent tests shall be submitted for the discussion of the concept. (2) Another schedule to be submitted at the same time shall contain type and number of those material specimens of the reactor pressure vessel which serve to determine changes in material properties resulting from operational influences caused by neutron irradiation after a certain time of operation, Surveillance specimens shall be used to assure the determination of strength and ductility properties of base material and welded joint following a certain neutron irradiation. (3) The schedules to be submitted in accordance with paras (1) and (2) above shall be agreed upon with the authorized inspection agency. (4) The pressure-retaining boundary shall be subjected to constant leakage monitoring. The localization of leaks shall be possible. (5) In areas of increased radiation levels the components to be inspected shall be provided with such insulations as will allow rapid removal and restoration if necessary. A mechanization of the ultrasonic examination shall be possible because of the radiation exposure of the personnel. (6) The inherent value and significance of the nondestructive examinations shall be confirmed by the authorized inspection agency prior to commissioning and on occasion of recurrent tests. (7) Settlement differences on systems of the pressureretaining boundary and on load-bearing structures shall be detectable. (8) The vibrational behavior of the components and systems of the pressure-retaining boundary shall be reviewed during commissioning and documented in test logs. Such review shall be carried out in the range between minimum and full load of the plant considering representative operational states. It shall be possible to carry out the corresponding measurements as recurrent tests and supplement them by measurements to be recorded during operation Destructive In-Process Tests (Surveillance Specimens) (1) For the irradiation program required in Subsec para. (2) the selection of materials and the establishment of other test conditions shall be agreed upon with the authorized inspection agency. In this context, the conditions contained in ASTM E shall be taken into consideration Recurrent Pressure Test (1) The recurrent pressure test shall be carried out in a way that will supply a safety-related result which is at least comparable to that of the initial pressure test. For this purpose, it shall be possible to carry out the recurrent pressure tests on the entire pressure-retaining boundary at the same pressure as used in the initial pressure test of the reactor pressure vessel. The testing temperature shall not be more than 55 C above the NDT temperature calculated for the time of testing. However, it need not be below 50 C. (2) As a rule, the testing intervals will be 8 years. (3) Following the recurrent pressure test, a nondestructive recurrent examination, e.g. an ultrasonic examination, 8

11 shall be carried out with regard to the reactor pressure vessel and at those points of the other components of the pressure-retaining boundary which are representative in view of stresses Recurrent Nondestructive Examinations of the Reactor Pressure Vessel Examination Technique (1) In the light of the present state of the art, the testing techniques to be employed shall be those of the ultrasonic examination. (2) The application of other testing techniques (such as eddy current, magnetic leakage flux, potential probe, penetrating dyes) may become necessary as an additional measure at certain locations if prior experience suggests that safety-related surface cracks will have to be anticipated. Such other testing techniques shall be used if it becomes obvious that the ultrasonic examination is not safe enough a tool for the detection of discontinuities. (3) The testing techniques and sound scanning directions shall be chosen in such a way that cracks and other discontinuities can be revealed with their possible orientations. Such orientations of discontinuities are: 1. the areas which are vertical to the main stress directions; 2. the fusion areas on welds; 3. the planes which are vertical to the direction of welds (transverse discontinuity detections). To examine the vessel wall for discontinuities which are not close to the surface, the tandem and single search unit techniques will usually have to be combined. (4) The sensitivity adjustment of the ultrasonic examination depends on the technique applied. The following provisions shall apply to some of the major techniques: 1. In case of a technique which uses a sound beam that hits the crack area at a right angle in the anticipated orientation, and in case of a tandem technique directed to this orientation, the recording limit shall at most be equal to the echo response of a circular reference reflector of 10 mm diameter with the orientations specified in para. (3), subparas 1, 2 and 3 above. This condition shall be complied with at any point of the ultrasonic beam cross section that has been employed. It shall be demonstrated by experiments that the usual fluctuations in the orientation of discontinuity areas are covered by this condition. 2. If, in case of a single search unit with a single crystal, only a diffuse scattering of crack areas is utilized for display purposes, i.e. if the sound intensity emitted at the blazing angle to the crack orientation does not arrive at the search unit, a circular reference reflector of 3 mm diameter having an orientation that is vertical to the sound beam shall be used as recording limit. This condition shall also be complied with at any point of the ultrasonic beam cross section that has been employed. (5) For other ultrasonic techniques such as scattering display technique, single search unit technique with separate transmitter and receiver crystals (transmitterreceiver unit), the required sensitivities shall be determined by experiments. (6) As a matter of principle, the sound intensity fluctuations resulting from coupling, attenuation and scattering shall be checked through transmission in the directions used for examination purposes. If this verification is not possible other methods such as the measuring of back scattering and the detection of shaperelated back reflections shall be used to carry out at least a coupling check. (7) The results of the verifications required in para. (6) above shall be recorded and taken into consideration for sensitivity calibrations and evaluations Required Examination Scope (1) As a matter of principle, there shall be a possibility of a visual inspection of the entire inner and outer surface of the wall of the reactor pressure vessel. Optical aids such as TV cameras may be used for this purpose. (2) For all areas of the reactor pressure vessel wall which do not permit visual inspection of the surface because of unavoidable design-related obstacles, an explanation shall be submitted of how possible defects such as gaping cracks, wear or detachment of weld attachments will be safely detected by other examination techniques. (3) For the inspection of the outer surface a sufficient gap with all necessary equipment shall be provided between reactor pressure vessel and biological shield. (4) After commissioning, the reactor pressure vessel, and in particular its welds, shall permit ultrasonic examinations over the entire volume of the wall with a sufficient detectability of discontinuities, either from inside and from outside as a complementation (or from outside and from inside as a complementation). (5) The design of the reactor pressure vessel, its surfaces, the conditions of its material including the condition of the weld claddings and the examination equipment shall be such that the entire volume of the reactor pressure vessel wall can be examined in a satisfactory way to the extent this is possible. Any restriction of the capability of examination shall be confined to the cases mentioned below provided the conditions specified in the following subparagraphs are complied with: 1. In volume areas of flanges which, as a result of the masking effect of coolant nozzle openings or bolt boreholes, cannot be reached with an ultrasonic beam in a suitable direction from the inner surface or the outer surface, a possibility of examination may only be waived if the following conditions have been complied with: All inner and outer zones which are close to the surface can be examined. The layer thickness to be examined is at least 35 mm. An examination for surface cracks can be performed. With regard to those volume zones inside the wall which cannot be examined it shall be demonstrated that only small stresses will be applied. If there are any welds in these zones it shall be demonstrated that the areas which cannot be examined are small in comparison with the critical crack sizes. 2. If certain bridges in borehole fields cannot be examined over their entire volume area in spite of the application of all possible ultrasonic techniques it shall be demonstrated that postulated cracks of the size of the areas which have not been examined cannot lead to such leaks or breaks as could not be coped with, in case of a sudden occurrence, by emergency core cooling and other actions of the reactor safety systems. In addition, the specified restrictions of the scope of examination on these bridges shall be taken into consideration when establishing the examination intervals (cf. Subsec ). (6) The ultrasonic equipment employed may be such that the search units for individual areas of the reactor pressure vessel are attached either only to the inner surface or only to the outer surface. It will have to be presupposed in this context that a sufficient examination of the entire wall cross section will be performed. In case this examination concept will have to be revised in the 9