Переводы документов. Translations in English

NP-010-16. Rules of design and operation of localizing safety systems for nuclear power plants

Approved by

Approved by

the Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated February 24, 2016  No. 70

 

FEDERAL RULES AND REGULATIONS

IN THE AREA OF ATOMIC ENERGY USE "RULES FOR ARRANGEMENT

AND OPERATION OF LOCALIZING SAFETY SYSTEMS

FOR NUCLEAR POWER PLANTS"

(NP-010-16)

 

I. Purpose and scope of application

 

1. These Federal Standards and Rules in the area of atomic energy use “Rules of design and operation of localizing safety systems for nuclear power plants” (hereinafter referred to as the Rules) are developed in accordance with Federal Law N 170-FZ “On the use of atomic energy” dated November 21, 1995 (Collected Acts of the Russian Federation N 48, 1995, art. 4552; 1997, N 7, art. 808; 2001, N 29, art. 2949; 2002, N 1, art. 2; N 13, art. 1180; 2003, N 46, art. 4436; 2004, N 35, art. 3607; 2006, N 52, art. 5498; 2007, N 7, art. 834; N 49, art. 6079; 2008, N 29, art. 3418; N 30, art. 3616; 2009, N 1, art. 17; N 52, art. 6450; 2011, N 29, art. 4281; N 30, art. 4590, art. 4596; N 45, art. 6333; N 48, art. 6732; N 49, art. 7025; 2012, N 26, art. 3446; 2013, N 27, art. 3451),  Decree of the Government of the Russian Federation No. 1511 “On approval of Regulation on development and approval of Federal standards and regulations in the field of atomic energy use” dated December 1, 1997 (Collected Acts of the Russian Federation, 1997, N 49, art. 5600; 1999, N 27, art. 3380; 2000, N 28, art. 2981; 2002, N 4, art. 325; N 44, art. 4392; 2003, N 40, art. 3899; 2005, N 23, art. 2278; 2006, N 50, art. 5346; 2007, N 14, art. 1692; N 46, art. 5583; 2008, N 15, art. 1549; 2012, N 51, art. 7203).

2. These Rules set the requirements for design, installation, operation of localizing safety systems for nuclear power plants including tests and repair.

3. These Rules apply to nuclear power plants that are under design, under construction, under operation and being decommissioned. Nuclear power plant units which are being decommissioned, are not subject to requirements of chapter III, as well as items 152 - 176, 190 - 192 hereof.

4. Procedure for bringing of nuclear power plants into compliance with these Rules, including terms and scope of required activities, is determined in each specific case in the construction, operation or decommissioning license conditions.

5. List of abbreviations is given in appendix No. 1 hereto, terms and definitions are given in appendix No. 2 hereto.

II. General requirements for localizing safety systems of

nuclear power plants

 

6. NPP unit shall provide for LSS ensuring performance of the following functions:

prevention or limitation of propagation of emitted radioactive substances out of the accident confinement area under normal operation and in case of abnormal, including accidents;

limitation of ionizing radiation discharge out of the accident confinement area under normal operation and in case of abnormal operation including accidents;

limitation of ambient pressure inside the containment in case of accidents;

reduction of concentration of radioactive substances discharged in the accident confinement area during accidents;

monitoring of concentration of explosion hazardous gases in hydrogen-containing mixtures in case of generation thereof in the accident confinement area under normal operation and abnormal operation, including accidents;

hydrogen explosion protection.

7. If in case of abnormal operation of NPP radioactive substances might be discharged beyond the room (container) in which they are contained, in the amount exceeding established limits for safe operation, accident confinement area borders shall be set for each relevant room (container), and LSS shall be designed to prevent or limit propagation of radioactive substances beyond the accident confinement area.

8. LSS and components thereof shall perform all functions provided for in the NPP design with due regard to influence of external natural and human induced factors which may arise in the NPP location area, as well as impacts occurring during accidents (e.g., shock waves, jets, flying items, forces from connected pipelines). Strength and operability of LSS components shall be substantiated for all normal operation modes, as well as for abnormal operation, including design-basis accidents.

9. LSS components operability under the influence of negative ambient temperatures able of resulting in water crystallization on their surfaces, shall be confirmed.

10. Protection of LSS components shall be ensured during NPP operation, against harmful impact of microorganisms and other biological objects.

11. LSS shall be able to perform their functions both when normal operation power supply sources are functioning as intended, and when they fail.

12. LSS components shall be accessible for examination, tests, repair, decontamination and maintenance.

For LSS components which are not accessible for examination and tests during NPP operation, the NPP design shall justify performance of design-basis functions by them during the designed operating lifetime of these components, or when performing safety analysis relevant NPP components shall be considered as inoperable.

13. Technical condition of NPP components, failure of which could influence operability of LSS and components hereof, shall be controlled with the frequency established in the NPP design documentation (hereinafter - NPP design) and in accordance with requirements of Federal standards and rules in the field of atomic energy use.

14. NPP design shall contain information on the number of tests which LSS and components hereof shall pass without loss of operability.

15. Reliability indicators calculations shall be performed for each LSS. LSS reliability indicators shall be considered when determining the value of probability of large accidental discharge.

16. All LSS active components shall be controlled from MCR, as well as (in case of MCR control failure) from ECR.

17. The volume of data presentation at MCR and ECR regarding the parameters under control, which describe operation of LSS and components hereof, the volume of recording and storage of relevant data, as well as requirements for redundancy of measuring channels and measurement accuracy shall be justified in the NPP design.

18. NPP premises containing liquid radioactive substances under normal operation or in case of abnormal operation, including design-basis accidents, shall provide for sealing steel lining of the floor and bottom part of walls. Wall lining shall be at least 200 mm higher than possible liquid level in case of full emptying of equipment or pipelines located in relevant room, as a result of destruction hereof.

19. Liquid level monitoring shall be provided for in premises with sealing lining of floor and walls, in case of accident with destruction of equipment or pipelines located in these premises, and displaying the information at MCR and ECR.

20. Design of LSS and components hereof, operation of them and documentation for LSS constitute the objects under quality assurance activities.

III. Requirements for arrangement of containment

 

General requirements

 

21. NPP power unit shall provide for the reactor facility containment. The need for containment for other systems (components) containing radioactive substances shall be justified in the NPP design.

The containment shall be able to perform the following functions under NPP normal operation and in case of abnormal operation, including accidents:

prevention or limitation of propagation of emitted radioactive substances out of the containment;

protection of personnel and population against ionizing radiation;

protection of systems and components enclosed in the containment, failure of which could result in discharge of radioactive substances out of the borders set by the design, in the amounts exceeding safe operation limits, against natural and man-induced impacts in cases stipulated by the NPP design.

22. NPP design shall set the list of NPP components, including building structures, forming part of the containment, inclusive as follows:

steel and reinforced concrete building structures, including ones with pre-stressing system and sealing lining;

items to be installed into the building structures of the  containment (penetrations, manholes, doors, gates, bypass and safety devices, as well as embedded parts for these components);

pipeline communications locations crossing the containment or connected hereto, within isolating devices and isolating devices themselves;

equipment and equipment communications going beyond the containment building structures and participating in formation of the accident confinement area.

23. Design-basis leakage value from the containment shall not be exceeded under NPP normal operation and in case of abnormal operation, including design-basis accidents.

24. NPP design shall envisage and justify engineering and organizational measures to limit containment leakage value in case of beyond design-basis accidents. Mentioned measures shall be aimed at limitation of ambient pressure and temperature within the accident confinement area, at prevention of explosive mixtures detonation, at protection of containment against dynamic impact of jets, flying objects, as well as at limitation of radioactive substances discharge to the environment.

In case of use of Corium Collecting and Cooling Device, reliable subcriticality of the medium enclosed herein, shall be provided.

Controlled discharge of radioactive substances out of the reactor facility containment is permitted in case of severe accidents only for prevention of containment destruction provided that the measures are taken to ensure radiation safety of population (by filtration of radioactive substances discharge, sheltering, evacuation or other measures).

25. The NPP design shall justify that the maximum value of overpressure (underpressure) in the environment enclosed in the containment, will not exceed design-basis pressure (underpressure) in case of design-basis accidents. Non-excess of design-basis temperature in case of design-basis accidents shall also be justified.

In cases when in order to prevent pressure rise inside the containment there are heat removal systems having active components (or passive components with moving parts), mentioned systems shall include several independent channels.

26. Design of LSS reinforced concrete structures shall be performed by observing the standards established in the Federal standards and rules in the field of use of atomic energy. The NPP design shall justify strength and serviceability of containment building structures.

27. The following requirements shall be observed for containments constructed as double-wall containment:

pressure value below atmospheric one, including the worst assumed wind conditions, shall be kept in the annulus under NPP normal operation and in case of abnormal operation, including design-basis accidents;

all leakages entering the annulus from the internal containment under NPP normal operation modes, in case of abnormal operation, including accidents, accompanied by pressure increase inside the internal containment, as well as in post-accident modes, shall be removed from the containment by ventilation system using filtration;

underpressure value control from MCR shall be provided for, as well as control of radioactive substances concentration in the annulus;

design-basis value of leakage to the environment shall not exceed 10% of design-basis value of leakage through the internal containment;

if there are pipelines in the annulus, rupture of which could result in pressure and (or) temperature increase in the mentioned space, it is required, that internal and external containments, as well as safety components located in the annulus, sustain relevant pressure and temperature loads;

the annulus shall form a uniform space for the purpose of mixing and dilution of radioactive substances discharge, entering with leakage from the internal containment in case of accident;

the annulus dimensions shall be sufficient to control technical condition and to repair containment building structures and to perform maintenance of equipment and pipelines located herein.

28. The NPP design shall justify strength of containment structures with due regard to the number of containment loading cycles during strength and leak-tightness tests for the whole designed NPP operating life, including commissioning and operational tests.

29. Containment building structures performing function of biological protection against ionizing radiation shall comply with requirements of sanitary-epidemiological rules and standards SanPin 2.6.1.24-03 “Sanitary rules for NPP design and operation", approved by Decree No. 69 of Chief Public Health Official of the Russian Federation, dated April 28, 2003 (registered by the Ministry of Justice of Russia under No. 4593, dated May 26, 2003). The measures for limitation of ionizing radiation discharge through gaps or welded joints between individual components of the containment shall be provided.

30. There are following requirements for containment reinforced concrete structures with pre-stressed reinforcement bars functioning without connection with concrete:

a possibility of control of the pre-stressing value shall be provided for each of the Containment Pre-stressing System components to be stressed, as well as a possibility for replacement hereof;

the NPP design shall justify possibility and necessary value for periodic restoration of tension of the stressed components, as well as conditions at which no stressed components tension restoration is allowed.

31. For containment, where overpressure could arise, the NPP design shall provide for the means of control and record of the Stressed-Strained States and temperature of containment building structures.

32. NPP unit operation with individual inoperable stressed components of the containment shall be justified in the NPP design.

33. Layout solutions taken in the NPP design shall prevent damage to building structures and safety-related components inside the containment, caused by pressure drop in case of abnormal operation, including accidents.

34. Design of isolating devices, manholes, doors, gates, safety and bypass devices of LSS shall provide a possibility of individual actuation and leak-tightness tests, as well as examination and repair with reactor shut down.

35. Thickness of protective layer providing corrosion protection of metal reinforcement bars of the containment reinforced concrete structures shall be determined during NPP design with due regard to aggressiveness of the environment and design-basis operating life of the structures.

 

Sealing lining

 

36. Reactor facility containment concrete surfaces shall have metal sealing lining.

37. Sealing lining parts joints between each other and other containment components shall allow regular leak-tightness examination. For steel sealing lining these joints shall be made by welding.

38. Methods, scope and frequency of the containment sealing lining welded joints examination shall be justified in the NPP design.

39. Rooms embodying any working media, which walls and floors form part of the containment, shall have the sealing lining made of stainless steel. Scope and frequency of the sealing lining welded joints examination shall be justified in the NPP design.

40. Type and distance of sealing lining anchoring shall be determined based on provision of the containment operability in case of accidents. Methods and locations of securing of sealing steel lining to the embedded parts of reinforced concrete structures of the containment shall be justified in the NPP design.

41. It is not allowed to weld scaffolding, stairs, cradles and other attachments used for installation, repair and maintenance of the containment, directly to the sealing lining.

Manholes, doors, gates

 

42. In order to exclude possibility of the containment depressurization during transportation of equipment and passage of personnel into the accident confinement area (and when leaving it) the containment shall be equipped with gates (locks).

43. Number of gates for reactor facility containment which are intended for passage of personnel shall be at least two.

44. Design of gates for personnel passage shall provide for mechanical interlock between doors (manholes), to prevent simultaneous opening of both ones. It is allowed to use electrical interlock for transport gate. Personnel gates (airlocks) shall be large enough to allow two men wearing anti-contamination clothing to move through while carrying a man on a stretcher.

45. Doors (manholes) shall be fitted with valves for equalization of pressure with indicators of their positions. Valves shall have interlock preventing simultaneous opening hereof on two doors (manholes).

46. Gate door (manhole) opening and closing mechanisms shall be equipped with electrical, hydraulic or other actuators. Mentioned mechanisms shall be actuated by a single man both outside, and inside the accident confinement area and gate.

47. Gates shall be equipped with MCR voice communication means.

48. Accident confinement area shall be accessed through gates intended for personnel passage from the controlled access area.

49. Connections of embedded parts (frames of manholes and doors, embedded parts for gates) with sealing lining shall be made by welding. Gate body connection with the embedded part shall also be welded.

50. If the NPP design provides for doors (manholes) for certain rooms inside the accident confinement area, and the NPP design provides for leak-tightness requirements for such doors (manholes), these also shall meet the requirements of present Rules.

51. Design of manholes, gates, doors and embedded parts hereof shall provide for observance of design requirements for leakage value, as well as for ionizing radiation dose rate attenuation both under normal operation and in case of abnormal operation, including design-basis accidents.

52. The NPP design shall justify, and NPP SAR shall specify the design-basis value of leakage occurring under design-basis pressure, through manholes, doors and gates, as well as through penetrations.

53. In the containment, where overpressure may arise, manholes, doors and gates shall open inside the accident confinement area. It is allowed to use structures, opening parts of which are moved in parallel to their aperture, provided that these are pressed by emergency overpressure to the frame from the accident confinement area.

54. Permissible values of doors (manholes) or gates closing time shall be justified in the NPP design and given in the NPP SAR.

55. Design of manholes, doors and gates representing containment components shall stipulate a possibility to monitor their leak-tightness from the outside in relation to an accident confinement area after each opening and closing cycle.

56. Position of manhole covers, door leaves, gate components representing containment components, shall be controlled at MCR.

57. Manholes used for evacuation, doors and gates shall be located above the maximum liquid level which may arise in the room in case of NPP abnormal operation, including design-basis accidents.

58. Personnel passage into the annulus shall be realized through the leak-tight doors installed in the external containment, which number shall be determined based on the fire safety requirements.  Doors shall be fitted with forced closing mechanisms (door assists).

 

Penetrations

 

59. Crossing of containment building structures by process and electrical communications shall be realized by using leak-tight penetrations.

60. Communications related to different safety system channels shall cross containment through different leak-tight penetrations.

61. Connection of leak-tight penetrations with the embedded parts and junction of the embedded parts with sealing steel lining shall be made by welding.

62. NPP design shall justify serviceability of leak-tight pipeline penetrations with due regard to impacts from connected pipelines, as well as serviceability of containment structures with due regard to thermal impacts from leak-tight penetrations.

63. Leak-tight penetrations shall be fitted with the inspection chamber for leak-tightness tests of welded joints. Permissible value of leakage through each penetration under design-basis working ambient pressure in the containment shall be established in the NPP design.

64. The NPP design shall provide for access to each pipeline penetration to enable performance of individual leak-tightness tests.

65. Leak-tight electrical penetrations design shall allow performance of individual check of penetration leak-tightness. Deviation from observance of this requirement shall be justified in the NPP design.

66. No leak-tight penetrations with gland sealing is allowed for application in the containment.

67. Measures for minimization of containment leak-tight penetrations number shall be taken during design.

Isolating devices

68. All pipelines crossing or connected to the containment shall be fitted with isolating devices to be installed as close as technically possible to the crossing location with containment. Number of isolating devices and installation locations shall be selected so that isolation of all pipelines crossing the containment, for which this is stipulated by the NPP design, shall be provided in case of any initiating event of design-basis accident and initiating event-independent failure of on safety system components considered in the NPP design.

69. Permissibility of failure of equipment with isolating devices of pipelines not connected to the reactor facility pipelines and equipment or atmosphere of the accident confinement area rooms and protected against potential external and internal impacts in case of accidents, shall be justified in the NPP design.

70. It is allowed not to install isolating devices on pipelines laid through the containment or connected hereto and used for intake of working medium from the primary circuit or accident confinement area rooms with further return to them, as well as for measurements during accident, provided that these pipelines and equipment connected by them meet requirements of these Rules for containment components.

71. Installation of pipeline hand-operated valves equipped with lock or welded plugs can be provided as isolating devices on pipelines used only during repair with reactor shut down.

72. If the need arises to close isolating devices, these shall be closed for such time that discharge of radioactive substances to the environment does not result in excess of design-basis limits and criteria of safety.

73. The NPP design shall specify conditions for each of active isolating devices at the border of accident confinement area, under which relevant isolating device closing signal shall be generated.

74. The NPP design shall justify for each isolating device the design-basis value of leakage through it beyond the accident confinement area border in closed position of the isolating device.

75. In case of compressed air pressure loss pneumatic isolating devices shall turn to position for performance of safety function.

76. Isolating devices being active components shall actuate automatically based on conditions determined in the NPP design.

77. The NPP design shall provide for measures to exclude unauthorized opening of isolating devices both during an accident, and in post-accident period, also in case of the drive power failure.

78. Pipeline valves used as isolating devices shall comply with the requirements of the Federal standards and rules in the field of atomic energy use, governing design, manufacture, tests, installation and operation of pipeline valves for nuclear power plants.

79. No check valves are allowed to be used as isolating devices.

 

Bypass and safety devices

 

80. Accident confinement areas where medium discharge is stipulated from one room to other or out of accident confinement area borders (in addition to discharge through passive steam condensers) in accordance with the NPP design for the purpose of containment destruction prevention, shall be fitted with safety and (or) bypass devices (e.g., discharge valves, burst discs) with filtration of medium discharged from the accident confinement area.

81. Containments not fitted with safety and (or) bypass devices shall be equipped with such for the period of containment strength tests.

82. Number of safety devices, flow capacity hereof, opening and closing pressure values shall be determined in the NPP design.

83. Safety and bypass devices of the containment shall be tested from time to time for actuation and leak-tightness with reactor shut down (in case of water coolant used in the primary circuit, reactor facility shall be cooled down). Frequency of tests and procedure hereof shall be justified in the NPP design and comply with Federal standards and rules in the field of atomic energy use.

Hydrogen explosion protection

84. Requirements of Federal standards and rules in the field of atomic energy use governing provision of hydrogen explosion protection at the nuclear power plant shall be observed for the reactor facility containment, as well as for systems, components and rooms located within the containment volume.

85. Using of materials in the reactor facility, equipment and building structures (including biological protection, heat insulation and corrosion protection coatings) not producing hydrogen under NPP normal operation and in case of abnormal operation, including accidents, is preferred in the NPP design.

List of potential hydrogen producing processes (sources) is given in appendix No. 3 hereto. Certain list of potential processes (sources) leading to hydrogen generation shall be justified in the NPP design.

86. The NPP design shall include analysis of generation, accumulation, distribution of hydrogen, criteria for determination of proper explosion protection in the accident confinement area, as well as justification of observance hereof under NPP normal operation and abnormal operation, including accidents.

87. The NPP design shall provide for the following:

composition of explosive hydrogen containing mixtures in the rooms within the space enclosed into the reactor facility containment, under normal operation, abnormal operation, including accidents;

locations and hardware for monitoring of hydrogen concentration, pressure and temperature of hydrogen-containing mixture in the rooms within reactor facility containment;

mechanical loads and temperature impacts to the reactor facility containment, caused by burning of hydrogen-containing mixtures in case of accidents, as well as possible consequences of mechanical and thermal impact of hydrogen-containing mixtures burning for NPP systems and components, including building structures within the reactor facility containment.

 

Hydrogen-containing Mixtures’ Parameters Monitoring

 

88. Hydrogen concentration within the reactor facility containment under NPP normal operation and abnormal operation including accidents, shall be monitored by at least two independent measuring channels. Data on hydrogen concentration shall be represented at MCR.

89. Number and location of hydrogen monitoring posts in the reactor facility containment shall be selected with due regard to possible hydrogen accumulation locations.

90. Data on hydrogen, water steam and oxygen concentration inside the reactor facility containment during an accident shall be represented at MCR and ECR, as well as at PERCP. Frequency of data acquisition shall be justified in the NPP design.

91. Hydrogen-containing mixture pressure and temperature measurement instrumentation shall be provided for within the reactor facility containment with displaying the information on these parameters at MCR and ECR, as well as at PERCP.

92. Signalization devices (acoustic and light) shall be provided for, represented at MCR and ECR, actuating in case of excess of the hydrogen concentration value in the accident confinement area as set in the NPP design.

IV. Requirements for design of pressure reduction systems, 

heat removal and medium treatment, and hydrogen explosion protection systems

 

Active sprinkler system

 

93. Active sprinkler system if it is provided for by the NPP design shall ensure performance of the following main functions:

ambient pressure reduction in the accident confinement area;

heat removal from the accident confinement area;

reduction of concentration of radioactive substances in the accident confinement area rooms.

94. Sprinkler system active components serviceability check shall be provided for during NPP unit power operation, including sprinkler pumps, with no system out of the availability mode.

95. Active sprinkler system shall be designed and made so that it could be checked in order to confirm compliance of consumption parameters with design requirements. Requirements for scope and frequency of sprinkler system tests shall be justified in the NPP design.

96. The NPP design shall provide for water drainage sumps for uninterrupted supply of active sprinkler system with working medium during accident and post-accident period. The following can be used as water drainage sumps: sump-tank, suppression pool, water drainage sumps of other safety systems, if the NPP design justifies that no combination of functions by components of mentioned systems results in failure to comply with NPP safety requirements.

97. The NPP design shall provide for measures to exclude non-homogeneity of solution in sprinkler system water drainage sumps, as well as means for treatment and keeping solution chemical composition, and measures to limit solution aggressiveness to materials inside the containment.

98. Water drainage sumps design shall be selected so to provide serviceability preservation of safety system channels using relevant water drainage sumps.

99. Water drainage sump design shall provide for treatment of water supplied to pumps, to remove mechanical contaminations, including from pipeline heat insulation washed from pipelines in case of rupture, and to exclude water loss at any NPP unit operation mode.

100. Water volume in the water drainage sump, design of filtering components hereof and intake devices shall provide simultaneous operation of all sprinkler pumps and pumps of other safety systems connected to that water drainage sump, with no failures of water supply to pumps with due regard to delay of water return to water drainage sump from accident confinement area rooms.

 

Ventilation and cooling systems

 

101. Ventilation and cooling systems shall provide performance of one or several of following functions:

heat removal from the accident confinement area;

creation of underpressure in the accident confinement area;

reduction of concentration of radioactive substances in the accident confinement area;

provision of necessary underpressure in the annulus (for double-wall containment of the NPP power unit).

102. Failure of the use of ventilation and cooling systems as part of LSS shall be justified in the NPP design.

103. Ventilation and cooling systems shall comply with requirements of Federal standards and rules in the field of atomic energy use, governing design and operation of nuclear power plants’ safety-related ventilation systems.

 

Passive steam condensation systems

 

104. Passive steam condensation systems, if these are provided for in the NPP design, shall ensure performance of following main functions:

ambient pressure reduction in the accident confinement area;

heat removal from the accident confinement area.

Passive steam condensation systems can be used for performance of other functions provided that there is justification in the NPP design.

105. Passive steam condensation systems shall have coolant margin providing condensation of steam generated in the design-basis volume in case of accidents with depressurization of pipelines inside the containment.

106. In case if walls of passive steam condenser form part of the containment, these are covered by requirements of chapter III thereof.

107. Pipelines, equipment, fixture components and other components of passive steam condensation system shall be designed for impact of steam and air mixture flow and other dynamic effects, arising of which is possible under NPP normal operation and in case of abnormal operation, including accidents.

108. The NPP design shall provide for measures to exclude damage to walls of passive steam condenser caused by hydraulic shocks which are possible during steam condensation, as well as by possible ambient underpressure in the accident confinement area.

109. The NPP design shall provide for measures to exclude non-homogeneity of solution in the volume of passive steam condensation system water drainage sumps.

110. The NPP design shall provide for measures to prevent reduction of capacity of passive steam condensation system due to accumulation of non-condensing gases.

111. The NPP design shall justify the value of time required for autonomous operation of passive steam condensation systems.

Hydrogen explosion protection systems

 

112. Hydrogen explosion protection systems shall еperform following functions:

prevention of generation of explosive mixtures within the ALA by maintaining hydrogen concentration in the mixture below the lower concentration limit for flame propagation;

prevention of explosion initiating source arising in the accident confinement area;

hydrogen concentration monitoring in the accident confinement area.

113. The following is applied as a part of hydrogen explosion protection systems:

hydrogen-containing mixtures combustion systems in the accident confinement area;

systems of hydrogen-containing mixtures removal from the accident confinement area, including purification and discharge hereof to the environment;

medium mixing systems in the accident confinement area;

accident and post-accident phlegmatization systems.

Composition and characteristics of systems providing hydrogen explosion protection shall be justified in the NPP design.

114. Hydrogen explosion protection systems shall perform their functions in case of inoperability of one (any one) active component or passive component with moving parts of relevant systems.

 

Emergency gas aerosol treatment plants

 

115. Emergency gas aerosol treatment plants, if these are provided for in the NPP design, shall ensure performance of following functions:

ambient pressure reduction in the accident confinement area;

reduction of concentration of radioactive substances in the atmosphere of accident confinement area rooms and in discharge of radioactive substances to the environment.

116. Filtering components of the emergency gas aerosol treatment plant shall be available for replacement under normal operation and during post-accident period.

117. Emergency gas aerosol treatment functions performance by ventilation and cooling systems or by hydrogen explosion protection systems is allowed provided that permissibility of combination of functions by mentioned systems is justified in the NPP design.

V. Seals

 

118. Seals of LSS components forming the border of accident confinement area shall provide leak-tightness required in accordance with NPP design under NPP normal operation, and in case of abnormal operation, including design-basis accidents.

119. Replacement of seals of LSS components (manholes, doors, gates, valves and other components) which can result in depressurization of the reactor facility containment, shall be performed with reactor shut down only (in case of water coolant used in the primary circuit, reactor facility shall be cooled down).

120. It is allowed to make sealing by welding with use of adapter components of individual doors, manholes, and repair ventilation systems’ communication components. Whereas, welded joints quality shall be checked, as well as compliance with requirements for LSS components, including leak-tightness requirements.

VI. Materials

 

121. Materials for production of LSS components shall be applied with due regard to conditions of operation hereof, physical, mechanical and process parameters in order to provide performance of functions by LSS components during the design-basis operating life.

122. Main materials in accordance with requirements of Federal standards and rules in the field of atomic energy use and welding materials in accordance with requirements of Federal standards and rules in the field of atomic energy use governing welding and buildup of equipment and pipelines of nuclear power facilities should be used for the equipment and pipelines which are covered by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities.

Main, welding and buildup materials given in Federal standards and rules in the field of atomic energy use governing general requirements for pipeline valves for the NPP are applied along with main materials, which application is permitted by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities, for manufacture of pipeline valves forming part of LSS.

123. Application of materials for the containment reinforced concrete structures is performed in accordance with requirements of Federal standards and rules in the field of atomic energy use governing design standards for LSS reinforced concrete structures.

124. Quality and properties of the basic materials for manufacturing, installation and repair of pressure-sealing steel linings, tanks and cases that are components of LSS shall comply with the requirements of standardization documents specifying the requirements for products, processes and other items of standardization in the area of atomic energy use as stipulated in par. 7 of the Regulations on standardization for the products (works, services) subject to requirements related to safety assurance in the area of atomic energy use as well as processes and other items of standardization associated with such products approved by Decree of the Government of the Russian Federation No. 669 dated June, 12, 2016 (Collected Acts of the Russian Federation, 2016, No. 29, art. 4839) and shall be confirmed by manufacturer's certificates.

125. In case of incompleteness of data in the certificate, the application of materials is allowed only after performance of necessary additional tests and researches confirming their complete compliance with the requirements of standardization documents. Results of performed additional tests and researches shall be made in the form of protocols (conclusions) to be attached to the original certificate of the materials manufacturing plant.

In case of delivery of a part of the scope of materials given in the certificate to other manufacturing plant, a copy of certificate shall be issued for the mentioned part of materials, bearing seal of the original certificate holder, indicating actual scope of delivered materials both in the copy and in the original certificate.

Original certificate shall be delivered in case of delivery of the whole scope (or a rest part) of materials.

In case of loss of the original certificate it is not subjected to re-issue, and the material can be applied for manufacture of LSS components only based upon the certificate duplicate issued by manufacturing plant of the material.

126. Welding materials which comply with requirements of Federal standards and rules in the field of atomic energy use governing main regulations for welding of LSS components, should be used in manufacture, installation and repair of sealing lining, tanks and covers related to LSS components.

127. It is required to perform tests of trial batch of materials and make report according to appendix No. 4 thereto, which report shall be approved by the head organization responsible for material study, for application of new materials in manufacture, installation and repair of sealing lining, tanks and covers representing components of LSS.

Possibility of application of new materials is realized in the form of a resolution agreed upon with the NPP design development organization and head organization responsible for materials study, which resolution is to be approved by the Operating organization.

Any new material shall be permitted for use subsequent to inclusion of the material standardization document into the Consolidated list of standardization documents as stipulated in par. 12 of the Regulations on standardization for the products (works, services) subject to requirements related to safety assurance in the area of atomic energy use as well as processes and other items of standardization associated with such products approved by Decree of the Government of the Russian Federation No. 669 dated June 12, 2016.

128. It is allowed to use new materials for manufacture, installation or repair of certain sealing lining, tank or cover representing LSS components, under the engineering solution prepared by engaging the head organization responsible for materials study and the NPP design development organization. Mentioned engineering solution with justifying materials confirming possibility of manufacture (installation, repair) by observing required quality, shall be submitted to the Operating organization. At the same time, scope and nomenclature of information to be submitted, as ones given in appendix No. 4 thereto, shall be determined by organizations made an engineering solution, depending on certain conditions of operation of sealing lining, tanks and covers.

Approved engineering solution shall be attached to the LSS certificate, and justifying materials shall be kept at the Operating organization during the whole LSS operation life.

129. Materials having different structural classes (perlite and austenite steel, non-ferrous metals) shall be transported and kept in conditions preventing contact between them.

130. Upon taking, materials shall pass incoming inspection (for mechanical, corrosion and other damages, for compliance of marking, quantity , weight and dimension parameters with data in the certificate), by recording the results of inspection, as well as inspection before discharge for production (installation).

131. Measures shall be provided for during manufacture and installation of LSS components (also, in case of transfer of marking, storage between operations) to exclude confusion of materials.

VII. Requirements for manufacture, installation and repair 

of sealing steel lining and embedded parts

 

132. Technology of manufacture, installation and repair sealing steel lining and embedded parts shall provide leak-tightness of the accident confinement area enclosure, in accordance with the NPP design, during the whole NPP operating lifetime. Mounting welded joints shall be accessible for inspection during operation. In case of inaccessibility of mounting junctions for inspection during operation, the NPP design shall justify performance of design-basis functions by these junctions during the designed operating lifetime.

133. Work performance plans for NPP shall provide for activities excluding damage to sealing lining and other LSS components during work performance on concrete pouring and equipment installation.

 

134. Manufacturing and installation documentation shall specify the list of communications laid through containment building structures or connected hereto, whereas the following shall be indicated for each communication:

connections of this with primary circuit pipelines, atmosphere inside the containment or components of other system located inside the containment;

name of system (component) connected with communication out of the containment and beyond the accident confinement area.

135. Anticorrosion coating of sealing steel lining of the containment shall be applied at the manufacturing plant. Anticorrosion coating of welded joints is performed after installation.

136. Measures shall be provided for in the accident confinement area to provide preservation of sealing lining (application of floor protection sheets, protective boards on walls, bumpers at possible shock locations during consignment transportation, preparation of locations, where focused loads could be applied during installation) during concrete pouring, installation of equipment and performance of other work.

137. Application of focused loads to sealing lining at points of slinging and supporting during storage and transportation, at locations of installation of heavy containment components and in other cases is allowed only based upon approval from the NPP design development organization.

138. Concrete pouring of flooring structures and walls, sealing lining of which is used as formwork, shall be performed layer by layer. Concrete pouring layer height, as well as sealing lining securing procedure shall be justified in the NPP design.

139. Welded joints of sealing lining components between each other and other containment components shall allow regular leak-tightness examination during installation, commissioning and operation. Containment components welded joints made in factory conditions are allowed not to be controlled individually during installation, commissioning and operation, if serviceability of relevant joints is confirmed during strength and leak-tightness integral tests of the containment.

VIII. Requirements for welding and control of welded joints

140. LSS components welding shall be performed according to processes developed in accordance with requirements of Federal standards and rules in the field of atomic energy use governing basic regulations for LSS components welding. Welding shall be performed in accordance with requirements of Federal standards and rules in the field of atomic energy use governing welding and buildup of equipment and pipelines of nuclear power facilities, in respect to the equipment and pipelines which are covered by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities.

141. LSS components welded joints quality control shall be performed in accordance with Federal standards and rules in the field of atomic energy use, governing NPP LSS welded joints control rules. Welded joint quality control shall be performed in accordance with requirements of Federal standards and rules in the field of atomic energy use governing welding and buildup of equipment and pipelines of nuclear power facilities, in respect to the equipment and pipelines which are covered by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities.

IX. Requirements for tests

General requirements

 

142. LSS and components and components hereof shall be checked in order to verify the compliance with the design parameters during commissioning, after repair, and periodically during an NPP power unit operating life.

143. Checkup of the localizing safety system and its components for compliance with design parameters shall be assured in performing the following tests:

strength test;

leak-tightness test;

functional test;

biological shielding test.

 

Depending on the designation of the localizing safety system, it shall be subjected to either to all mentioned tests or their separate types in accordance with the requirements of these Rules and the NPP design.

144. Testing of the LSS components for compliance with the design characteristics subsequent to their manufacturing shall be performed by the manufacturing plant in accordance with testing programs approved in due course as established at the manufacturing plant.

145. The NPP design shall provide tests methodologies for LSSs and their components after installation (construction), during pre-commissioning and operation. Tests of LSSs and their components shall be carried out according to programs developed by the operating organization on the basis of the NPP design documentation.

146. Measures shall be provided in the course of testing to prevent damage of the containment components as well as other  safety-related components of the NPP in case of any deviation from the test parameters stipulated in the NPP design.

147. Leak-tightness and strength tests of the containment shall be performed after completion of installation of all containment components, as well as auxiliary and control systems in the scope required for performance of all functions by the containment according to the NPP design.

148. Test medium supply and discharge lines shall be provided for test of containment under overpressure. Discharge line in strength tests of the containment, as well as leak-tightness tests under design-basis pressure shall be equipped with safety devices.

149. Protocols, statements and certificates shall be made based upon results of tests of LSS and components hereof, recommended samples for which are given in appendix No. 5 hereto. Test results shall be recorded into the LSS certificate to be made in accordance with item 209 hereof.

150. Defects revealed during tests shall be eliminated, and tests repeated.

151. It is prohibited to test containment for strength and leak-tightness with failed safety devices.

 

Strength tests of containment

 

152. Strength tests of the containment are performed under overpressure, as well as with underpressure once per the whole NPP power unit operating life, during pre-commissioning of the NPP power unit.

The NPP design shall determine criteria of the need for repeated strength tests. Resolution on performance of repeated strength tests upon reaching specified criteria shall be taken by the Operating organization.

Test medium overpressure value in strength tests of the containment shall be equal to the design-basis pressure increased 1.15 times.

154. The following is required during containment strength tests:

experimental determination of dynamics of change of the Stressed-Strained States parameters at checkpoints determined in the NPP design;

comparison of test data with design-basis ones and (or) limit permissible vales determined in the NPP design.

For the purpose of determination of stress values at checkpoints and comparison hereof with design-basis ones, overpressure or underpressure values in strength tests should be changed at the rate and with keeping time as determined in the NPP design for relevant pressure values (steps).

Further change of working medium pressure to the next value (step) of test pressure can be performed only after determination of reliable conclusion on compliance of the containment with design-basis strength criteria.

 

156. Following parameters shall be recorded during strength tests of the containment:

data of visual examination of accessible external surfaces of the containment;

Stressed-Strained States parameters of the containment at checkpoints;

containment components temperature;

stresses in the reinforcement ropes of the Containment Pre-stressing System, on which jacking force monitoring sensors are installed;

medium parameters in the volume of accident confinement area;

ambient temperature out of the containment;

data of geodetic surveillance of containment components displacement.

These parameters (except for ambient temperature) should be measured at checkpoints of the containment, which shall be specified in the NPP design and in test program.

157. Containment strength assessment criteria shall be justified in the NPP design. Criteria of assessment of Stressed-Strained States shall be values or changes of them given in item 156 hereof for parameters with relevant test pressure value. These criteria shall be given in the containment test program developed on the basis of the NPP design.

 

Leak-tightness tests of containment

 

158. Containment leak-tightness testing by air pressure for verification of compliance with the design pressure shall be performed once in the period of pre-commissioning activities (after completion of construction and assembly works before the initial loading of fuel to the reactor), after that the testing shall be performed at least once in 10 years or with frequency justified in the NPP design, as well as after repair or replacement of components which have impact on leak-tightness or strength in case these components cannot be monitored locally.

159. Containment leak-tightness test during operation of the NPP shall be performed with reactor shut down under reduced (below the design-basis value) overpressure (in case of water coolant used in the primary circuit, reactor facility shall be cooled down). Reduced pressure value, as well as frequency of containment leak-tightness tests with reduced pressure shall be justified in the NPP design and presented in the NPP SAR.

160. Containment (or autonomous parts hereof) for which NPP design provides for keeping of underpressure under normal operation and in case of abnormal operation, including accidents, shall be tested periodically for leak-tightness with underpressure during NPP operation. Test pressure value, as well as frequency of containment underperssure tests during operation shall be justified in the NPP design and presented in the NPP SAR.

161. For NPP power units which LSS include bubbler and vacuuming system, tests shall be performed with design-basis and (or) reduced air pressure, to confirm performance of functions of the part of the containment, which serves as air trap, as well as with design-basis underpressure for the part of the containment, where underpressure is created during accidents.

162. Working medium levels in each water drainage sump and tank during containment leak-tightness tests shall comply with working levels under normal operation.

163. Qualified methodologies shall be used for measurement and calculation of leakage from containment and its individual components during leak-tightness tests.

164. Revealed defects, for example, leakage locations, shall be documented as duly established by the Operating organization.

Equipment inside the containment shall sustain tests without failures in operability.

 

165. In the process of reactor facility containment integral leak-tightness tests medium parameters in ALA (pressure, temperature, humidity) should be registered at least once an hour until the following criterion is satisfied:

 

base_1_214422_32768 at base_1_214422_32769,

 

where base_1_214422_32770 is tolerance of leakage value determination, % per day;

L is leakage value obtained during tests, % per day;

base_1_214422_32771 is confidence probability.

166. Containment leak-tightness tests during pre-commissioning shall be performed at least with two pressure values, reduced and design-basis ones. Whereas both pressure steps shall be kept during the period of parameters stabilization in the accident confinement area (in accordance with requirements of item 165 hereof).

167. Assessment criteria for results of containment tests with design-basis pressure shall be the value of leakage given in the NPP design. Whereas the following inequation is to be met:

 

base_1_214422_32772,

 

where L is the leakage value obtained in the course of testing in compliance with the requirements set in par. 166 of these Rules, % per day; Ld is the leakage value specified in the NPP design, % per day; base_1_214422_32773 is the leakage estimation error, % per day.

168. Assessment criteria for results of containment leak-tightness tests during operation is compliance with the following inequality:

 

Lc < 1.15 Lcr,

 

where base_1_214422_32774 is the leakage value obtained during tests in the course of operation, % per day;  Lcr is the leakage value obtained during pre-operational tests with reduced pressure in the course of commissioning, % per day.

169. The leakage value obtained during pre-operation testing with reduced pressure in the course of commissioning (Lcr) shall be recorded in the containment certificate in order to be used as the criterion for testing within the NPP operation period.

170. The obtained value of containment leakage shall be attributed to average value of pressure in the accident confinement area during containment leak-tightness tests for the period of recording.

171. Containment medium parameters at which integral tests are performed shall be justified in the NPP design.

172. The rate of increasing and decreasing pressure in ALA during the leak-tightness test with due regard to tolerance of determination hereof shall not exceed values justified in the NPP design.

173. The NPP design shall provide possibility of medium discharge from the accident confinement area through filters after containment leak-tightness tests during operation.

174. If absolute method of determination of the leakage value is applied for the containment leak-tightness test, mentioned tests shall be performed in accordance with basic requirements for measurements during integral tests of the containment, given in appendix No. 6 thereto.

Leak-tightness tests

of containment components

 

175. Leak-tightness tests of containment components (manholes, gates, isolating devices, tight doors and penetrations) during NPP power unit construction and commissioning shall be performed stage by stage, as far as containment construction mounting work is completed. Containment components shall be accessible for performance of mentioned tests.

Containment components subjected to leak-tightness tests shall be determined in the NPP design.

176. Gate seals shall be tested after each opening and closing cycle during power operation of the NPP power unit, as well as before startup of the NPP power unit.

 

Hydraulic tests for leak-tightness of rooms, 

water discharge sumps and tanks

 

177. Rooms mentioned in items 19, 39 thereof, water discharge sumps, as well as tanks shall be subjected to hydraulic tests. NPP design shall provide for filling in and drainage systems for mentioned rooms.

178. Hydraulic tests shall be performed during commissioning of rooms, water drainage sumps and tanks and from time to time during the whole NPP power unit operating life (frequency shall be determined in the NPP design), as well as after performance of repair affecting sealing components.

179. In case of the need for tests in winter, measures shall be taken to prevent water freezing. Hydraulic tests should be performed at temperature of supplied air of 5 °С and more. Use of salt solutions is prohibited.

180. As far as rooms, water drainage sumps and tanks are filled in with water, structures shall be under supervision as well as leakages from control cavities. In case of leakage detection, test shall be stopped, water discharged and leakage cause eliminated.

181. Room, water discharge sump and tank shall be considered as passed hydraulic tests, if no leakages arise on the tank wall surface or by edges of bottom (for rooms and water drainage sumps - from control cavities) during 24 hours, as well as no water level reduction is detected, except for water level reduction caused by change of liquid temperature during tests.

 

Functional tests of localizing safety systems 

and components hereof

 

182. The NPP design shall provide for performance of functional tests of LSS and components hereof during preparation of the NPP unit for commissioning and from time to time during NPP operation.

183. LSS and components hereof during operation shall be tested with the frequency justified in the NPP design, and with due regard to requirements of Federal standards and rules in the field of atomic energy use.

184. Functional tests of LSS and components hereof during NPP operation shall not result in them out of availability.

185. Compliance of following basic parameters and indicators with the NPP design shall be confirmed during tests of LSS components:

during commissioning of the NPP power unit:

flow rate characteristics of pumps and gas blowers of systems;

capability of filtering components to perform their functions;

operability of instruments;

capability of pumps to perform their functions at minimum permissible water level in the water discharge sump;

time from the pump startup beginning to water ingress to the water discharge sump beginning;

water level in the water discharge sump;

functional capability of pipeline valves;

design-basis parameters of sprinkler injectors.

from time to time during operation:

characteristics of pumps and gas blowers of systems;

operability of instruments;

functional capability of pipeline valves.

Pipelines and injectors of sprinkler system shall be checked by air for the flow capacity with reactor shut down at frequency justified in the NPP design.

187. It is necessary to check functional capability of isolating devices, including determination of the time value required for closing hereof, during functional tests of isolating devices.

188. Functional tests of LSS active components shall be performed with simultaneous check of serviceability of their normal actuators.

189. Active isolating devices shall be subjected to functional tests with reactor shut down at frequency justified in the NPP design, except for cases, when the need for and safety of checking of individual parameters of isolating devices with power operation of reactor is justified in the NPP design.

 

Tests of biological shielding of localizing 

safety systems (components)

 

190. LSS (LSS components) biological shielding tests shall be performed during commissioning of the NPP power unit.

191. The following containment locations are subjected to biological shielding test:

locations of doors, manholes, gates and penetrations;

locations of possible presence of personnel (under normal operation, during and after accidents) at the containment external side.

Scope of tests, certain test locations, as well as design-basis dose rates of ionizing radiation shall be specified in the NPP design.

Biological shielding structures are considered appropriate for operation, if ionizing dose rate at locations determined in the NPP design does not exceed design-basis values.

 

X. Tests of biological shielding of localizing safety systems

and components hereof

 

General requirements

 

193. Administrative management of the NPP shall be obliged to use LSS and components hereof in accordance with the requirements of the Rules, as well as to assign a person responsible for supervision of LSS and components hereof, and the person responsible for operable condition hereof, from the engineers and technicians passed knowledge checking as duly established by the Operating organization.

194. Maintenance and repair of LSS and components hereof shall be performed by trained persons, who are permitted to unsupervised work with relevant systems and components, as duly established by the Operating organization.

195. LSS shall be installed and tested in full and ready for performance of functions stipulated by the NPP design before delivery of fuel to the NPP.

196. No NPP power unit startup, as well as power operation of the reactor facility are allowed in the following cases:

with value of leakage from the containment exceeding the design-basis value of leakage;

with inoperable LSS components, as well as components of auxiliary and controlling safety systems required for normal functioning of the NPP, in the amount which fails to comply with established conditions of safe operation;

with failed reinforcement ropes of the containment in the amount, at which no safe operation conditions determined in the NPP design are observed;

with failed bypass and safety devices of the containment;

with leakage of working medium from LSS water discharge sumps, value of which exceeds value justified in the NPP design.

 

197. Safe operation conditions shall be determined in the NPP SAR and given in the NPP power unit operation process regulations for LSS and components hereof, including limitations for deviations of process parameters of the LSS and components hereof, requirements for minimum composition of operable LSS components, as well as conditions for putting out of operation of LSS components for maintenance, repair and tests.

198. Requirements of operation guideline of LSS and components hereof, as well as requirements of the NPP power unit operation process regulations shall be observed during operation of LSS and components hereof.

199. Technical condition monitoring shall be performed for keeping LSS and components hereof operable, including with regard to corrosion wear, maintenance, repair, tests and checks. The NPP administrative management shall develop LSS inspection program establishing requirements for the scope and frequency of inspection of technical condition of LSS and components hereof, maintenance, repair, tests and checks based on design requirements, requirements of NPP power unit operation process regulations, as well as with due regard to results of previous inspection. The results of the inspections and tests shall be documented in the form of certificate and recorded to the LSS certificate.

200. Tests of LSS and components hereof during operation are performed in accordance with requirements of chapter IX hereof.

201. After repair LSS components shall be tested for compliance with design-basis parameters.

202. No repair is allowed at pressurized equipment and pipelines of LSS.

203. Position and functional capability of isolating devices shall be checked before putting of LSS and components hereof into operation after repair, as well as after long-term reactor facility shut down (over three days).

204. After completion of repair and check of repaired component or the LSS as a whole, an enumeration of the repair work and test results are to be entered in the certificate.

205. In NPP conditions with reactor shut down (in case of water coolant used in the primary circuit, reactor facility shall also be cooled down), reactor facility containment depressurization is allowed provided that special engineering and organizational measures are taken, which sufficiency shall be justified in the NPP design. In other cases, when radioactive substances present inside the containment in amounts capable result in accident if out of the containment, containment depressurization is prohibited.

206. The need and procedure of personnel access for maintenance during operation of the NPP power unit in the accident confinement area with power operation of the reactor, as well as with hot reactor facility, shall be justified in the NPP design and reflected in the NPP power unit operation process regulations.

Documentation requirements

207. The NPP design shall provide for documentation requirements, in accordance with which LSS components are manufactured and installed.

208. Operation guidline for LSS and components hereof shall specify the scope and frequency of maintenance and checks of operability of LSS and components hereof, determined with due regard to requirements of these Rules, NPP design and results of tests during pre-commissioning.

209. The certificate shall be made by the Operating organization for each LSS.

LSS certificate shall contain the following information:

LSS name;

list of components (equipment, pipelines and building structures) of LSS;

LSS test program;

information on the person responsible for LSS operable condition;

information on LSS pipelines, as follows:

purpose, operating life;

safety class of pipeline, category of seismic resistance;

year of installation completion, identification of the pipeline drawing and name of the manufacturing plant of parts and assembly units of the pipeline, installation organization name, information on working medium, design-basis and working pressure and design-basis temperature;

information on pipeline materials;

values of nominal outside diameter, thickness of walls and length of pipes;

information on the revealed defects and repair of them;

information of containment building structures (if any in the LSS) and other building structures, namely:

data on reinforced concrete structures (grade and density of concrete, compression strength grade of concrete, data on frost resistance and waterproofness of concrete, data on the type of reinforcement structures - stressed/non-stressed, quantity and design of reinforcement structures);

safety class of containment (other building structures), seismic category;

materials and design of sealing lining;

materials and design of embedded parts;

number of pipeline and cable leak-tight penetrations of various type, design of leak-tight penetrations, applied materials;

number of gates, information about design and materials of gates;

information on Stressed-Strained States of the containment, obtained during strength tests of the containment;

data on forces in stressed reinforcement tendons of the containment at sensor locations;

information on chemical composition and mechanical parameters of welded joint materials and chemical composition of buildups;

value of leakage (design-basis and actual) from the containment in the period of pre-commissioning under design-basis pressure;

value of leakage (actual) from the containment in the period of pre-commissioning under reduced pressure;

parameters and results (obtained values of leakages) of local and integral tests performed during operation of the NPP;

information on the revealed defects and repair of them;

information on LSS equipment by each unit, namely:

name of equipment, purpose, operating life;

name of the manufacturer, factory identification number and year of manufacturing;

safety class, category of seismic resistance;

technical characteristics of equipment and main parts hereof, including information about working medium, design-basis pressure and temperature;

 

information on chemical composition and mechanical parameters of parts, welded joints and buildups (only technical composition for the latter);

information on the thermal treatment;

information on non-destructive testing results for parts, welded joints and buildups, revealed defects and repair of them;

test parameters and results;

information on preservation and packaging;

conclusion on compliance of manufactured equipment with Federal standards and rules in the field of atomic energy use and engineering documentation;

warranty obligations.

The following shall be attached to the LSS certificate:

certificate of manufacture for parts and assembly units of pipelines;

pipeline installation certificates;

LSS equipment and pipelines certificates;

containment components manufacture certificates;

installation certificates for containment and other LSS components;

operation guideline containing information necessary for checking of basic dimensions and compliance of equipment with requirements established by the NPP design, as well as information about fitting with valves if it is supplied along with equipment;

copies of compliance certificates for LSS components, compliance of which is subjected to confirmation in the form of mandatory certification;

calculations of strength or extracts from them with reference to calculations and description of input data and results.

210. No information on equipment and pipelines which are covered by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities, is included into the LSS certificate.

211. During operation the LSS certificate shall be supplemented with data on performed tests, technical examination, in-service inspection of metal condition, repair, replacement, as well as values of operating life characteristics determined during operation.

 

operating life control and extension

 

212. The Operating organization shall develop the program for control of operating life of equipment, pipelines and building structures of LSS.

213. Residual operating life shall be confirmed at periodic safety assessment of the NPP.

214. In case of extension of operating life of the equipment, pipelines or building structures of LSS, the Operating organization, if there is residual operating life, shall perform justification of possibility of further operation of mentioned equipment, pipeline or building structures based on results of the program of operating life control in accordance with requirements of Federal standards and rules in the field of atomic energy use governing operating life control.

 

 

 

 

 

XI. Technical examination and registration of LSS 

and components hereof

215. LSS and components hereof shall undergo technical examination before commissioning and from time to time during operation. Technical examination of equipment and pipelines which are covered by Federal standards and rules in the field of atomic energy use governing requirements for design and safe operation of equipment and pipelines of nuclear power facilities, should be performed in accordance with requirements of mentioned Federal standards and rules.

216. The purpose of technical examination is to determine that LSS components are manufactured and mounted in accordance with requirements of these Rules and design documentation, are operable and could be used in the following period.

217. Technical examination includes:

verification of documentation;

external and internal visual examination of LSS components, including supports, in accessible locations;

integral tests (during technical examination of the containment);

local (individual) tests (if there is instruction in the NPP design on the need to perform local tests);

presentation of results.

Locations inaccessible for examination because of radiation conditions shall be determined by the Operating organization. External examination inaccessibility for other reasons shall be determined by the NPP design development organization and the Operating organization.

218. The Operating organization shall make the list of LSS components, which are not accessible for internal and (or) external examinations because of structural peculiarities, process reasons (e.g., impossibility of emptying) or radiation situation. In each particular case for such LSS components, the Operating organization shall develop technical examination guidelines.

The list shall be approved by responsible state regulatory authority for safe use of atomic energy.

219. Technical examination is subdivided as follows:

primary technical examination, to be performed before the beginning of commissioning, after replacement, modernization or capital repair of LSS components;

regular technical examination, to be performed during NPP operation;

extraordinary technical examination, to be performed in following cases:

after dynamic impacts having human induced or natural origin, which intensity complies with design-basis values of exceeds them;

in case of failures of NPP normal operation which lead to change of LSS components operation parameters above values determined in the design documentation;

according to resolution of the Operating organization or upon the request from the state authority regulating safe use of atomic energy.

220. The following is performed during technical examination:

check of availability of design and engineering documentation (only during primary technical examination);

check of LSS certificate;

analysis of documentation which contains results of pre-in-service inspection of metal condition (only during primary technical examination);

analysis of documentation which contains results of previous in-service inspection of metal condition;

analysis of results of examination of LSS components condition and testing hereof, performed in accordance with item 217 hereof.

221. The following is performed during examinations of LSS components:

check of readiness of equipment, including components hereof, pipelines, parts and assembly units, and building structures, for commissioning and operation (check is performed during primary and extraordinary technical examination);

visual inspection to reveal superficial defects, including mechanical and corrosion damage, and erosion failures;

assessment of condition of supports, suspensions, fixtures and detachable junctions.

222. Primary technical examination of equipment and pipelines having protective housings and covers, shall be performed before welding hereof.

223. Technical examination of LSS (LSS components) during NPP operation is performed in accordance with schedules to be developed by the NPP administrative management based on the requirements established in the NPP design documentation and presented in NPP SAR.

224. Technical examination is performed by the operating organization. At least 10 days prior the Operating organization shall notify the responsible state regulatory authority for safe use of atomic energy about the readiness of LSS for technical examination, place and date of the technical examination.

225. Before the technical examination LSS equipment shall be emptied from the working medium, and surfaces subjected to inspection shall be cleaned from dirt.

226. LSS equipment and pipelines in contact with radioactive media, shall be decontaminated before the beginning of technical examination.

227. In case of revelation of defects during performance of technical examination, the defect unit examination act shall be made and sent to the NPP design development organizations, to manufacturing plant and to the responsible state regulatory authority for safe use of atomic energy.

228. Results of technical examination shall be recorded in the technical examination certificates, with protocols of integral and local tests attached. Based on these certificates, the Operating organization shall take resolution on results of the technical examination, indicating permissible conditions of operation and dates of next technical examinations, with making relevant entries in LSS certificates.

229. After primary technical examination of LSS (LSS components), the registration shall be performed according to requirements of Federal standards and rules in the field of atomic energy use governing rules of compliance assessment of safety-related products in the field of atomic energy use.

No. LSS (LSS components) registered before entering into force of these Rules are subjected to re-registration.

 

 

 

 

 

 

 

 

 

 

 

 

Appendix No.1

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

DESIGNATIONS AND ABBREVIATIONS

 

NPP       -              Nuclear Power Plant

MCR      -              Main Control Room

Containment     -              Containment / leak-tight enclosure

PERCP   -              Protected Emergency Response Control Post

LSS         -              Localizing Safety System

NPP SAR              -              Nuclear Power Plant Unit Safety Analysis Report

PCW      -              precommissioning works

ECR        -              Emergency Control Room

CPSS      -              containment prestressing system


 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 


Appendix No. 2

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

TERMS AND DEFINITIONS

 

Explosion protection (hydrogen) - engineering and organizational measures providing prevention of detonation of hydrogen-containing mixtures in the reactor facility equipment and in the space enclosed in the reactor facility containment, as well attenuation of impact of burning of hydrogen-containing mixtures to the reactor facility containment and other safety-related systems and components of NPP, under normal operation of the NPP, and in case of abnormal operation, including accidents.

Containment - the aggregate of components, including building structures which, while enclosing a space around the reactor facility or other radioactive substances containing facility, form the border as provided by the design and prevent propagation of amounts of radioactive substances and ionizing radiation in the environment exceeding pre-determined limits.

Leak-tightness - capability of an component or a system to limit propagation of liquid and gaseous substances, including aerosols, beyond the established border.

Design-basis pressure (for the containment) - the value of medium overpressure within the accident confinement area determined by the design development (engineering) organization and providing the containment strength as well as non-exceedance of the design-basis leakage from the containment.

Leakage value - a quantitative characteristic of leak-tightness to be determined as mass of volume of medium discharged from the controlled space at given parameters per unit of time.

Design-basis value of leakage is the value of leakage specified for the system (component) in the NPP design.

Actual value of leakage is the value of leakage obtained during checks (tests) of the system (component).

Accident confinement area - a space enclosed into the containment (or other LSS components) within which retention of radioactive substances emitted during an accident is provided for.

Isolating device - pipeline valves designed for isolation of accident confinement area from the environment.

Integral tests - testing of containment for leak-tightness or strength, which results confirm design-basis characteristics hereof (non-excess by actual value of leakage of the design-basis leakage value or a capability of the containment to sustain loads determined in the NPP design for design-basis accidents).

Local (individual) tests - tests of localizing safety systems components, which results confirm design-basis value of leak-tightness of relevant components.

Localizing safety systems (components) -systems intended for prevention or limitation of propagation of radioactive substances and ionizing radiation out of the boundaries established by the design as well as discharge of them to the environment.

Manhole, door - an component of the containment providing passage of persons and (or) transportation of equipment through building structures enclosing the accident confinement area.

Lower flame propagation concentration limit - the minimum content of combustible material in homogeneous mixture with oxidizing medium, at which flame propagation is possible in the mixture to any distance from the inflammation source.

Penetration (leak-tight) - an component of the containment providing crossing of building structures enclosing the accident confinement area (by preserving leak-tightness of the containment), by pipelines, air conduits, power cables, ionization chamber channels, etc.

Containment depressurization - a condition of the NPP components at the boundary of the accident confinement area, which will lead to excess of the design-basis leakage value from the containment in case of arising of the design-basis accident initiating event with design-basis functioning of safety systems.

Underpressure (for the containment) - the value of ambient vacuummetric pressure within the accident confinement area, to be determined by the design development (engineering) organization, at which containment strength is provided.

Design-basis temperature - the value of medium temperature within the accident confinement area or temperature of the containment components determined by the design (engineering) organization and providing for the containment operability.

Lock -  a facility (room) or a device representing the containment component and intended for access of people and (or) transportation of consignment into/out of the accident confinement area with preservation of the containment leak-tightness.

 

Appendix No. 3

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

LIST

OF POTENTIAL PROCESSES (SOURCES) LEADING TO

HYDROGEN GENERATION

 

1. Under normal operation:

VVER type reactors:

corrosion of structural materials;

water radiolysis in the primary circuit (particularly in the nuclear core);

water radiolysis in the spent fuel pool;

discharge of hydrogen contained in the coolant, during repair process associated with loss of integrity of the primary circuit;

for RBMK, EGP type reactors:

water radiolysis in the repeated forced circulation circuit (in main circulation circuit), including in the nuclear core, as well as in the reactor control and protection system circuit at reactor power operation and at shutdown;

water radiolysis in spent fuel pools;

corrosion of structural materials;

BN type reactors:

in process equipment assemblies, where water is used as coolant;

spillage or leakage of sodium and further contact with water of it;

equipment cleaning from sodium;

water radiolysis in the spent fuel pool and other systems where water might be exposed to intensive irradiation;

when sodium interacts with oil in normal operation conditions, and also with spirits or other liquids used for washing.

 

2. In case of accidents:

water and water steam radiolysis in the core (for VVER and RBMK types);

water and steam radiolysis outside the core;

discharge of hydrogen and oxygen contained in the reactor facility circuits before the accident;

decomposition of ammonia contained in the coolant (for VVER);

decomposition of hydrazine and ammonia dispensed into the makeup water at compensated leak (for VVER type reactors);

decomposition of hydrazine contained in tanks and dispensed into safety systems from the tanks at accidents (for VVER type reactors);

thermal dissociation of water (at temperatures exceeding 2000 °С);

zirconium-steam reaction;

iron-steam reaction;

steam-graphite reaction (for RBMK reactor types);

depressurization of process pipelines for feeding of hydrogen;

corrosion of aluminum or zinc materials of heat insulation and structural components (for VVER and RBMK type reactors);

corrosion of carbonic steel with damaged organic coating in boric acid solution (for VVER type reactors);

interaction of uranium dioxide with water steam;

contact of sodium with concrete (for BN type reactors);

nuclear fuel contact with concrete.

 

 

Appendix No. 4

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

REQUIREMENTS

FOR THE REPORT JUSTIFYING APPLICATION OF NEW MATERIAL

 

1. In order to include new materials into the list of documents for standardization of main materials used in manufacture, installation and repair of sealing lining, tanks and covers related to LSS components, the following shall be specified in the report:

a) purpose of the material;

b) chemical composition information;

c) type and method of production of semi-finished products;

d) value of limit temperature Tmax, at which use of material is allowed;

e) information on working media where use of materials is allowed;

f) information on the thermal treatment;

g) values of yield point, ultimate resistance, relative elongation and contraction;

h) values of elasticity modulus, Poisson ratio, linear expansion ratio, heat conductivity ratio and material density;

j) characteristics of resistance to brittle failure;

k) characteristics of change of cyclic strength;

l) corrosion resistance characteristics.

2. The characteristics specified in subparagraph "g", paragraph 1, Appendix No. 4 shall be determined within the temperature range from 20 °C to Tmax with the interval of 50 °C as well as at the temperatures exceeding  Tmax by 25 °C and 50 °C.

3. Quantitative data characterizing time-to-time variations of the parameters described in subparagraph "g", paragraph 1, Appendix No. 4 shall be presented.

4. The characteristics specified in subparagraph "h", paragraph 1, Appendix No. 4 shall be determined within the temperature range from 20 °C to Tmax with the interval of 100 °C as well as at the temperature exceeding  Tmax by 50 °C.

 

 

Appendix No. 5

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

(recommended sample)

REPORTS, STATEMENTS AND CERTIFICATES

ON THE TESTING RESULTS FOR THE CONTAINMENT

AND ITS COMPONENTS

                                 TEST REPORT

               for the containment

 

      _______________________________________________________________

      (preliminary, after completion of construction, regular)

            __________________________________________________

                    (as a whole or any independent part thereof)

            __________________________________________________

                       (leak-tightness, strength)

 

                                  Unit No. ____ of _____________ nuclear power plant

 

                                                    ______________ 20__

 

1. On the results of testing ________________________________________________

                           (preliminary, after completion of construction,

                                            regular)

 of the containment ___________________________________________________

                                (as a whole or any independent part thereof)

for leak-tightness.

1.1. The tests were performed in accordance with the requirements of par. No. __________ of the testing program

within the period from ________ till ___________

 

    Air pressure variation schedule for the accident confinement area,

parameter record sheets for determination of leakage values as well as the list of

detected containment defects are attached to this

report.

 

 

 

1.2. Leakage values are determined for the following number of tests:

___________________________________________________________________________

 

Values of test air pressure inside the containment

and results of calculations are given below.

 

 

Leakage value and absolute error of its measurement, %/ per day

Confidence probability

Initial test pressure, kPa

Beginning of test at specified pressure

Date

Time, h

 

 

 

 

 

 

 

 

 

 

 

1.3.   Obtained leakage values are compared (in accordance with

the requirements of par. No. ____   of the testing program) with the leak-tightness criteria

and comply (fail to comply) with the specified requirements.

 

2. On the results of testing ________________________________________________

                           (preliminary, after completion of construction,

                                            regular)

 of the containment ___________________________________________________

                                (as a whole or any independent part thereof)

for strength.

 

2.1. Tests were performed according to the requirements of par. No. _______________

of the working program within the period from _______________ till ____________________

(see par. 1.1 hereof).

 

    Parameter recording sheets as well as the list of detected

containment defects are attached to this report.

 

 

 

2.2. Stressed-strained state of the containment

___________________________________________________________________________

                    (as a whole or any independent part thereof)

is determined for ____________ of test air pressure values in the accident

confinement area equal to _____________ kPa.

 

    Assessment of stressed-strained state was based on the

readings of _________ transducers with concurrent visual inspection of the concrete surface

in order to detect any cracks (in accordance with the requirements of par. No. _________

of the working program).

 

    Stress values in the reinforcement under the test pressure of ___________ kPa

did not exceed _____________ kPa. Except for the zones ________ where stresses

up to _________ kPa were registered.

 

    Cracks with the width of _________ mm are were detected

at the elevations of ____________ in the zones ___________.

 

    After pressure reduction in the containment cracks  ______________

(closed, unclosed)

 

2.3. Measured values of stresses, strains (movements), inclinations,

as well as the recorded width of cracks ____________ the design values.

                                         (do not exceed,

                                           exceed)

 

                                CONCLUSION

 

    Containment ____________________________________________

                                  (as a whole or an independent part hereof)

of Unit No. _______________________________ of _____________________________ nuclear power plant:

_______________________________

   (passed, failed) the leak-tightness tests;

_______________________________

   (passed, failed) the strength tests

 

Head of the Testing

Commission ________________________________

                                                  (signature, surname)

 

Members of the Commission ____________________________________

                                                  (signatures, surnames)

 

 

 

                                 TEST REPORT

       for recording of parameters in the course of leak-tightness tests

                             of containment components

 

      _______________________________________________________________

      (preliminary, after completion of construction, regular)

                 ________________________________________

                    (as a whole or any independent part thereof)

 

                                  Unit No. ____ of _____________ nuclear power plant

 

                                                    ______________ 20__

 

 

 

 

 

 


Date of testing

Measurement time, h, min

Pressure inside the containment, kPa

Mass average temperature inside the containment, °С

Mass average gas constant inside the containment, J/(kg·°C)

Time from the beginning of tests, h, min

Note

gauge

barometric

absolute

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Persons in charge        ______________________________

                                                         (signatures, surnames)

 

 

 

                                 LIST

 

of defects detected during tests of the containment __________________________________

                                        (preliminary, after construction

___________________________________________________________________________

                       completion, regular)

___________________________________________________________________________

                    (as a whole or any independent part thereof)

___________________________________________________________________________

                       (leak-tightness, strength)

 

                                  Unit No. ____ of _____________ nuclear power plant

 

                                                    ______________ 20__

 

Date and time for detection of defects (leaks) ____________________________

Detection group (team) ________________________________________________

Head ______________________________________________________________

                                       (surname)

 

Defect (leakage) searching route  _________________________________

                                                    (Location No.)

Additional information on the route _____________________________________

                                              (elevation)

 

Test conditions

Location of defects (leaks)

Marking of defects

Detailed description of defects

Note

defect number

date of testing

 

 

 

 

 

 

 

 

 

 

 

 

 

Persons in charge        ______________________________

                                                         (signatures, surnames)

 

                               

 

 

 

                                      TEST REPORT

            for recording of parameters in the course of strength tests

                          of the containment

 

      _______________________________________________________________

             (preliminary, after completion of construction,

                              post-repair)

                 ________________________________________

                    (as a whole or any independent part thereof)

 

                                  Unit No. ____ of _____________ nuclear power plant

 

                                                    ______________ 20__

 

Test beginning date

Test beginning time, h, min

Test pressure inside the containment, kPa

Humidity inside the containment

Pressure transducer location inside the containment

Pressure transducer

Counting of time from the test beginning, s

Measured temperature value inside the containment, °С

Increment of measured temperature value inside the containment, °С

Note

Elevation

Section line

Number

Type

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Persons in charge        ______________________________

                                                         (signatures, surnames)




                             CERTIFICATE

 

of elimination of defects detected in the course of testing ______________________

                                                      (preliminary,

___________________________________________________________________________

               after completion of construction, post-repair)

 of the containment ___________________________________________________

                                (as a whole or any independent part thereof)

___________________________________________________________________________

                       (leak-tightness, strength)

 

                                  Unit No. ____ of _____________ nuclear power plant

 

                                                    ______________ 20__

 

1. Defects specified in the following Lists of Detected Defects were eliminated:

No. ____ dated __________ to Test Report ___________ No. ____ dated __________

 

2. All marked defects ____________________________________________________

                                      (eliminated, not eliminated)

___________________________________________________________________________

                   (if no specify the defect marking

                  and the cause for its non-elimination)

 

Repair works were performed by the team headed by: ____________________

___________________________________________________________________________

                                 (surname)

 

3. Control of the repair works was carried out by the following method _______________

 

4. Control results _______________________________________________________

 

Persons in charge          ______________________________

                                                     (signatures, surnames)

 

Person in charge from the special-purpose

acceptance department                  ______________________________

                                                     (signature, surname)

 

Acceptance inspector in charge   ______________________________

                                                     (signature, surname)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Appendix No. 6

to federal  rules and regulations

in the Area of Atomic Energy Use

"Rules for arrangement

and operation of localizing safety systems

at nuclear power plants”,

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

No. 70 dated 24.02.2016

 

GENERAL REQUIREMENTS

FOR MEASUREMENTS IN THE COURSE OF INTEGRAL TESTING OF THE CONTAINMENT

BY THE "ABSOLUTE" METHOD

 

1. Air pumped into the accident confinement area shall have:

relative humidity not exceeding 15% at the ambient temperature if absolute testing pressure does not exceed 0.5 MPa;

relative humidity not exceeding 25% at the ambient temperature if absolute testing pressure does not exceed 0.25 MPa;

relative humidity not exceeding 30% at the ambient temperature if absolute testing pressure does not exceed 0.17 MPa;

relative humidity not exceeding 40% at the ambient temperature if absolute testing pressure does not exceed 0.15 MPa.

2. Air pumped into the accident confinement area shall not contain any admixture of oil and dust above 0.002  g/m3 and 0.01 g/m3 respectively.

3. The automatic measurement system for the parameters shall assure measurements of local pressure, temperature and air humidity values in various points of the accident confinement area with pre-defined tolerance.

4. Pressure measurements shall be provided at least in three different points inside the containment, the above-mentioned measurements shall be independent from each other. Average pressure value is determined by two of them in the relevant set of measurements, and the third measurement is redundant, its value shall be displayed at the compressor plant control panel.

Pressure gages used for pressure measurement in determination of leakages should meet the following requirements (irrespective of the anticipated leakage value):

for the pressure measurement range - (0 - 1.15 Pd) MPa where Pd is the design pressure;

for underpressure measurement range - (0 - 0.06) MPa;

for accuracy class – at least 1.5.

Barometric pressure gages should meet the following requirements:

for measurement range - (0.09 - 0.11) MPa;

for accuracy class - 0.3.

The data of the local meteorological station are acceptable as barometric pressure values.

6. The following requirements shall be satisfied in order to provide representativeness of the mass average temperature:

no temperature transducers are installed in rooms with the volume of less than 200 m3;

one temperature transducer shall be installed in rooms with the volume of 200 - 700 m3;

in rooms with the height exceeding 5 m at least two temperature transducers shall be installed (one transducer per each 5 m of height);

in rooms with the volume exceeding 700 m3 temperature transducers shall be installed on the basis of one transducer per 700 m3 with spacing of 5 m along the room height.

7. Transducers for measurement of the ambient temperature inside the containment during determination of leakage values shall meet the following requirements depending on the expected temperature value:

for measurement range: 0 - 100 °C;

for measurement error: not exceeding 0.1 °C.

 

8. In order to determine the value of average moisture content in the working medium inside the containment the humidity content transducers shall be installed at the points with the highest expected gradients of temperature defined in the design documentation.

9. Humidity content transducers shall be installed inside the containment on the basis of one transducer per each 10000 m3, but at least one.

10. Transducers used for measurement of humidity in the atmosphere inside the containment in the course of leakage detection shall meet the following requirements:

for measurement of dew point - according to par. 7 hereof;

relative humidity measurement range shall be 0 - 100%;

absolute error of measurements shall not exceed 3%.

11. Calculation and statistical processing of hourly leakage values shall be performed in order to monitor and analyze the test progress.

 


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