Approved by

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

FEDERAL RULES AND REGULATIONS

IN THE FIELD OF NUCLEAR ENERGY USE "BASIC

REQUIREMENTS FOR JUSTIFICATION OF STRENGTH OF

VVER TYPE REACTOR INTERNALS"

(NP-102-17)

I. Purpose and scope

1. These Federal rules and regulations in the area of nuclear energy use "Basic requirements for justification of strength of VVER type reactor internals" (NP-102-17) (hereinafter referred to as Basic Requirements) have been developed in accordance with Article 6 of the Federal Law No. 170-FZ "On Nuclear Energy Use" dated November 21 and the Decree of the Government of the Russian Federation No. 1511 "On approval of the Regulation on development and approval of Federal rules and regulations in the area of nuclear energy use" dated December 1, 1997 (Collected Legislation of the Russian Federation, 1997, No. 49, art. 5600; 2012, No. 51, art. 7203).

2. These Basic Requirements shall be applied by the designer of reactor facility, as well as by the operating and leading materials technology organization during justification of strength of reactor internals for nuclear power plants with VVER type reactor facilities (designed, under construction and in operation).

3. The provisions of these Basic Requirements do not apply to justification of strength of reactor internals in the following cases:

beyond design basis accidents;

damage of reactor internals during transportation and process operations.

4. The list of abbreviations is presented in Appendix No. 1, terms and definitions in Appendix No. 2, conventions in Appendix No. 3 hereto.

II. General

5. Justification of strength of internals shall be provided in the reactor facility project, and the results of justification shall be presented in SAR for the NPP. In case of modifications of the design of internals, application of new construction materials, changes in standards of water chemistry regime of primary coolant of the reactor facility and changes in design-basis operating conditions, justification of strength of internals shall be performed taking into account these changes.

6. Justification of strength of internals shall be based on the results of calculations confirming that the limit states of the internals, taking into account the safety factors, will not be achieved throughout the design life in all operation modes of the internals provided for by the reactor facility project.

7. The methods used for justification of the strength of internals during operation shall take account of all loads on the internals during operation and allow to establish numerical values of factors determining achievement or non-achievement of the limit states during operation.

III. Requirements for physical and mechanical

characteristics of construction materials of

reactor internals

8. Construction materials for manufacture of internals shall be selected taking into account their radiation and corrosion resistance in the primary coolant medium of VVER type reactors. Physical and mechanical characteristics of these materials shall ensure the designated operating life of internals.

9. Physical and mechanical characteristics of the materials of internals shall be defined within the temperature range covering all operation modes of the internals provided for by the design, taking into account the impact of exposure. The range of radiation doses and temperatures at which the physical and mechanical characteristics of the materials are determined shall be sufficient for justification of strength of internals throughout the designated operating life. Operation of components of internals within dose and temperature ranges where the physical and mechanical characteristics have not been determined, or their predicted values have not been justified, is not allowed.

10. All physical and mechanical (modulus of elasticity, strength limit, yield limit, true tensile stress, Poisson's ratio, uniform elongation, elongation, contraction, strain hardening, characteristics of radiation-induced creep and radiation-induced swelling, characteristics of resistance to corrosion cracking, parameters of cracking resistance, fatigue curves, dependences determining crack growth rate during operation) and thermophysical (thermal conductivity, thermal capacity, thermal linear expansion coefficient) characteristics of the materials required for justification of strength of internals shall be determined by the leading materials technology organization. Numerical values of the specified physical and mechanical characteristics shall be reflected in the documents on standardization in the area of nuclear energy use included in the summary list of documents on standardization provided for by clause 6 of the Regulations on standardization concerning products (works, services) for which the requirements related to safety assurance in the area of nuclear energy use, as well as processes and other objects of standardization related to such products are established, approved by the Decree of the Government of the Russian Federation No. 669 dated July 12, 2016 (hereinafter – the Summary List) (Collected Legislation of the Russian  Federation, 2016, No. 29, art. 4839).

11. The leading materials technology organization shall ensure storage of the documents on the basis of which the numerical values of physical, mechanical and thermophysical characteristics of materials are established until the end of the operating life of internals.

12. The characteristics of cracking resistance and dependences determining crack initiation and growth rate during operation used during justification of strength of internals shall take account of degradation of materials during operation and shall be determined on the basis of experimental studies, taking into account the radiation dose, radiation-induced swelling, radiation-induced creep, temperature, chemistry of the coolant and the results of tests of materials of internals in research reactors.

Characteristics of radiation-induced swelling shall be determined on the basis of experimental and theoretical studies, taking into account the impact of radiation dose and temperature, exposure rate, state of stress and the results of tests of materials of internals in research reactors.

Characteristics of radiation-induced creep shall be determined on the  basis of experimental and theoretical studies, taking into account the  impact of radiation dose, exposure rate, radiation-induced swelling and the results of tests of materials of internals in research  reactors.

Characteristics of strength and ductility shall be determined on the basis of experimental and theoretical studies, taking into account the impact of radiation dose and temperature, radiation-induced swelling, state of stress and the results of tests of materials of internals in research reactors.

IV. Criteria of strength of reactor internals

13. The conditions of failure to achieve the following limit states during operation shall be used as criteria of strength of internals:

a) plastic deformation throughout any cross-section area of a component of internals;

b) origination of a crack <1>;

--------------------------------

<1> This state is used as a limit state at the stage of design.

c) unstable development of a crack;

d) loss of stability (global and/or local);

e) ultimate plastic deformation which, when reached, may lead to destruction;

f) unacceptable change in geometrical dimensions.

14. Numerical values of strength criteria for the limit state specified in subclauses b, c, and e of clause 13 of these Basic Requirements shall be determined by the leading materials technology organization, and for the limit state specified in subclause f of clause 13 – by the designer of internals. Justification of the specified numerical values of criteria shall be provided in the reactor facility project and included in SAR for the NPP. In case of modifications of the design or operating conditions of internals, application of new materials, changes in standards of water chemistry regime of primary coolant of the reactor facility, justification of new numerical values of the specified strength criteria shall be performed, or their previous values shall be confirmed.

V. Safety factors for strength criteria

of reactor internals

15. The justifications of strength of internals shall incorporate safety factors for all limit states listed in clause 13 of these Basic Requirements shall be provided for. Numerical values of safety factors shall ensure failure to achieve the limit states during operation, taking into account the calculation error, and must be justified experimentally and/or confirmed by the experience in operation of VVER type reactors (except for the safety factors established by these Basic Requirements).

16. During justification of numerical values of safety factors, the following must be taken into account:

a) experience in operation of internals of similar design (if any);

b) the results of experiments at test facilities and in research reactors;

c) calculation error for radiation damage of components of internals, parameters of stress and strain state and fracture mechanics;

d) error due to a spread of values of physical and mechanical characteristics of materials.

VI. Requirements for design justification of strength

of reactor internals

17. Calculations of strength of reactor internals shall contain justification of failure to achieve the limit states listed in clause 13 of these Basic Requirements by all structural components of internals during operation.

18. The following shall be taken into account during calculation:

a) the impact of primary coolant of the reactor facility, neutron irradiation and temperature on the physical and mechanical and thermophysical properties of materials of internals;

b) radiation-induced creep and radiation-induced swelling of materials of internals during operation;

c) neutron flux and temperature distribution throughout the height and thickness of components of internals and in azimuthal direction;

d) changes in geometrical dimensions and shape of internals during operation (if any);

e) mechanical loads due to the nuclear core weight, own weight of internals, contraction forces of hold-down assemblies and fuel assemblies;

f) hydraulic loads;

g) coolant pressure difference;

h) Archimede's (buoyant) force;

i) dynamic loads due to falling of control devices of the control and protection system in case of emergency shutdown;

j) dynamic loads due to external dynamic impacts;

k) forces due to interaction between components of internals and components of the nuclear core (in case of contact);

l) forces due to interaction between components of internals and other reactor components.

19. The strength of components of internals shall be justified by the following calculations:

a) static strength;

b) stability (global and/or local);

c) initiation of a fatigue crack, taking into account the impact of the coolant;

d) initiation of a crack by a corrosion cracking mechanism.

e) formation of UEZ of a material;

f) stable development of a crack;

g) unstable development of a crack (quasi-brittle fracture);

h) unacceptable change in shape;

i) ultimate plastic deformation;

j) vibration strength;

k) external DI.

20. Stress and strain state of internals shall be determined according to the sequence of operation modes established in the reactor facility project, taking into account mechanical and temperature loads, changes in physical and mechanical characteristics of materials during operation, as well as radiation-induced swelling and radiation-induces creep.

VII. Calculation of static strength

21. Calculation of static strength shall be performed for justification of failure to achieve the limit state specified in subclause a of clause 13 of these Basic Requirements during operation for all components of internals and all operation modes of reactor facility established by the design.

22. During calculation of static strength, all mechanical and hydraulic loads on internals shall be taken into account. Stresses in components of internals shall be classified according to Appendix No. 4 to these Basic Requirements. Residual stresses shall not be taken into account during calculation of static strength.

23. Calculation of static strength of internals shall be based on limitations of size of stress category groups presented in Appendix No. 4 to these Basic Requirements, relative to the value of nominal allowable stress [] at the design temperature. Hardening of materials of internals due to irradiation shall not be taken into account during calculation of static strength of internals.

24. The nominal allowable stress value for components of internals (except for fasteners) shall be taken equal to the smallest of the following values:

.

For fasteners of internals, the nominal allowable stress value shall be determined as per the formula:

.

25. Safety factors shall be at least 2.6 for the strength limit n_m and at least 1.5 for the yield limit n_0.2.

26. During determination of nominal allowable stress values, R_m^T and R_p0.2^T values shall be established in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List.

27. The total value of stresses included in the group of stress categories ()₁ presented in Appendix No. 4 to these Basic Requirements shall not exceed:

[] – in case of NO;

1.2 [] – in case of AO;

1.4 [] – in case of DBA.

28. The total value of stresses included in the group of stress categories  ()₂ presented in Appendix No. 4 to these Basic Requirements shall not exceed:

1.3 [] – in case of NO;

1.6 [] – in case of AO;

1.8 [] – in case of DBA.

For fasteners of internals, average tension stresses due to mechanical loads in the cross-section shall not exceed []_w.

Average tension stresses due to mechanical loads and temperature impacts shall not exceed:

1.3 []_w – in case of NO;

1.6 []_w – in case of AO;

1.8 []_w – in case of DBA.

For fasteners of internals, total tension stresses due to mechanical and temperature loads causing tension, bending and torsion, shall not exceed:

1.7 []_w – in case of NO;

2.0 []_w – in case of AO;

2.4 []_w – in case of DBA.

30. For all components of internals and all NO modes provided for by the reactor facility project, average bearing stresses ()_s due to mechanical load shall not exceed 1.5R_p0.2^T, and average shear stresses ()_s shall not exceed 0.5 [] (in fasteners – 0.25 R_p0.2^T).

In case of joint impact of mechanical and temperature load, average shear stresses ()_s shall not exceed 0.65 [] (in fasteners – 0.32 R_p0.2^T).

VIII. Calculation of stability

31. Calculation of stability shall be performed for justification of failure to achieve the limit state specified in subclause d of clause 13 of these Basic Requirements during operation for all components of internals and all operation modes established by the reactor facility project.

32. Calculation of stability shall be performed in presence of loads causing compression and/or shear stresses.

33. During calculation of stability of internals, the loads achievement of which will cause a global loss in stability of internals or a local loss in stability of structural components of internals, both under static and dynamic loads (critical loads), shall be determined. Justification of the fact that all loads that may result in a loss in stability during operation for all structural components of internals will not exceed their critical values, taking into account the safety factor of 2, shall be provided based on the results of calculations.

34. During calculation of stability of internals, the following shall be taken into account:

a) changes in the dimensions and shape of internals;

b) possible mechanical interaction between components of internals and components of the nuclear core;

c) the results of experiments (if any) for determination of the critical values of loads for internals.

35. Numerical values of critical loads or critical stresses for structural components of internals shall be determined using software or analytical dependences, subject to justification of the possibility of their use for structural components of internals.

IX. Calculation of a fatigue crack initiation

36. Calculation of a fatigue crack initiation shall be performed in order to establish the time of crack initiation in case of a cyclic loading for all components of internals in NO and AO modes provided for by the reactor facility project.

37. The conditions of a fatigue crack initiation shall be established based on the design (provided for in the reactor facility project) or actual (during operation) sequence of operation modes of internals, according to fatigue curves, taking into account the impact of the coolant, residual stresses, irradiation, radiation-induced swelling and vibration loads.

38. Calculation of a fatigue crack initiation in components of internals shall be performed in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List.

X. Calculation of a corrosion crack initiation

39. Calculation of a corrosion crack initiation shall be performed in order to establish the time of crack initiation in case of a static and cyclic loading by a corrosion cracking mechanism for components of internals contacting with the coolant in NO and AO modes provided for by the reactor facility project.

40. The conditions of corrosion cracking initiation shall be established taking into account the residual stresses, radiation-induced swelling, radiation-induced creep and loading history of components of internals.

41. Calculation of corrosion cracking in components of internals shall be performed in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List.

XI. Calculation of formation of

ultimate embrittlement zone of a material

42. Calculation of formation of UEZ shall be performed for determination of size of the zone in a structural component of internals where a metal has achieved the ultimate embrittlement state due to radiation exposure. The calculation shall be performed for all components of internals subject to radiation exposure in NO and AO modes provided for by the reactor facility project.

43. UEZ shall be considered as a geometrical zone in a structural component of internals where radiation-induced swelling exceeds the critical value established by the leading materials technology organization.

44. From the moment of formation of UEZ in a structural component of internals, a postulated crack with the dimensions determined by the dimensions of UEZ shall be defined in this component.

45. Calculation of formation of UEZ and determination of dimensions of a postulated crack shall be performed in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List.

XII. Calculation of stable development of a crack

46. Calculation of stable development of a crack shall be performed for determination of maximum permissible dimensions of cracks in structural components of internals during operation of internals.

47. Calculation of an increase in dimensions of a crack in NO and AO modes shall be performed both for design (initiated and postulated) cracks and defects detected during control of the state of metal of internals.

48. Calculation of an increase in dimensions of cracks shall be performed in cases when the time to formation of UEZ and/or initiation of a corrosion or fatigue crack in a component of internals is less than the design operating life of reactor facility, as well as in presence of process defects.

49. Calculation of an increase in dimensions of a crack contacting with the coolant shall be performed taking into account the fatigue, creep and corrosion cracking.

50. Calculation of an increase in dimensions of a crack not contacting with  the coolant shall be performed taking into account the fatigue and creep.

51. During calculation of an increase in dimensions of a crack by a fatigue mechanism, vibration loads on components of internals shall be taken into account.

52. During formation of UEZ, a calculation of stable development of a crack by fatigue, creep and corrosion cracking mechanisms shall be performed, taking into account the increase in dimensions of UEZ due to irradiation during operation of reactor facility.

53. Calculation of stable development of a crack shall be performed in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List.

XIII. Calculation of unstable development of a crack

54. Calculation of unstable development of a crack shall contain justification of failure to achieve the limit state specified in subclause c of clause 13 of these Basic Requirements during operation of the reactor, up to the end of its operating life, for all components of internals and all operation modes of reactor facility established by the design.

55. The analysis of unstable development of a crack shall be performed taking into account the results of calculations of stable development of a crack (Chapter II of these Basic Requirements).

56. Calculation of unstable development of a crack shall be performed in accordance with the requirements of documents on standardization in the area of nuclear energy use included in the Summary List of documents on standardization.

XIV. Calculation of unacceptable change in

geometrical dimensions

57. Calculation of unacceptable change in geometrical dimensions (including due to radiation-induced swelling and thermo-irradiation creep) shall contain justification of failure to achieve the limit state specified in subclause f of clause 13 of these Basic Requirements for all structural components of internals by the end of their operating life.

58. During calculation of unacceptable change in geometrical dimensions of structural components of internals, all mechanisms of deformation of materials of internals causing irreversible changes in geometrical dimensions, including plastic deformation, radiation-induced swelling and radiation-induced creep, shall be taken into account.

59. In NO, AO and DBA modes, changes in geometrical dimensions of components of internals, as well as the gaps between the internals and components of the nuclear core, shall be within the limits established by the design. Changes in geometrical dimensions of components of internals shall not prevent normal functioning of control devices of the reactor protection system.

60. In case of DBA due to main coolant pipeline rupture the structure of reactor internals in only subject to the requirements for limitation of movement preventing normal functioning of control devices of the reactor protection system, cooling and subsequent disassembling of the nuclear core.

61. Calculation of unacceptable change in geometrical dimensions of structural components of internals shall be performed using software intended for calculations of deformation of structures due to plastic deformation, radiation-induced swelling and thermo-irradiation creep.

XV. Calculation of ultimate plastic deformation

62. Calculation of ultimate plastic deformation of structural components of internals shall be performed for justification of failure to achieve the limit state specified in the subclause e of clause 18 of these Basic Requirements during operation of reactor facility for all structural components of internals and all operation modes of the internals provided for by the reactor facility project, including dynamic impacts.

63. Calculation of ultimate plastic deformation of structural components of internals shall be performed using software verified for calculations of structures under plastic deformation conditions.

XVI. Calculation of external dynamic impacts

64. Calculation of external dynamic impacts shall contain justification of failure to achieve the limit states specified in clause 13 of these Basic Requirements under dynamic loads transmitted to internals through the reactor vessel for all components of internals and all operation modes of the internals established by the reactor facility project.

65. The values of dynamic loads on components of internals shall be determined by calculation of DI for reactor facility performed in accordance with the requirements of Federal rules and regulations in the area of nuclear energy use, taking into account the simultaneous impact in two horizontal directions and vertical direction, based on the data of the NPP designer.

66. Calculation of external DI for internals shall be performed by dynamic analysis method (by accelerograms) or linear spectral method (by response spectra), using SW. When using the linear spectral method of calculation, the absence of physical and geometrical nonlinearity in case of deformation of structures of internals shall be demonstrated.

67. Application of static method of calculation of seismic impacts on internals is only allowed in cases when the lowest frequency of natural oscillations of structural components of internals is above 20 Hz. At that, if this frequency is within the range of 20-33 Hz, the overload coefficient for operating loads of 1.3 shall be set.

68. The value of relative damping k_D shall be determined on the basis of experimental studies; in the absence of studies, the value of k_D shall be taken equal to 0.02.

69. In case of external DI, combinations of design loads and allowable stress values for structural components of internals shall be specified in accordance with Appendix No. 5 to these Basic Requirements.

XVII. Calculation of vibration strength

70. Calculation of vibration strength of internals shall be performed according to the results of the analysis of vibrations detected during commissioning works at VVER type reactors, using the results of measurements of vibration characteristics obtained at test facilities.

71. Calculation of vibration strength of internals shall be performed for justification of the fact that resonance oscillations with unacceptable displacement amplitudes of structural components of internals due to vibration loads caused by the coolant flow and/or its pressure oscillations will not be observed in all operation modes of the internals provided for by the reactor facility project. Characteristics of natural oscillations of structural components of internals (oscillation frequencies and decrements) shall ensure that vibration stresses if the internals will not exceed the values established in the reactor facility project by more than 15% (without regard to vibrations) in case of all disturbances.

72. Calculation of vibration strength of internals shall include the following:

a) determination of natural frequencies and modes of oscillations of components of internals (calculation of natural frequencies of oscillations of structural elements shall be performed using SW);

b) check of the absence of vibration shock interactions of components of internals with each other and the nuclear core components, in order to avoid excessive wear;

c) calculation for a fatigue crack initiation, taking into account vibration stresses in accordance with Section IX of these Basic Requirements.

73. During calculations of vibration strength of internals, databases of vibration characteristics measurement results obtained at test facilities and during measurements at VVER type reactors in the course of commissioning works shall be used.

Appendix 1

to federal rules and regulations

in the field of nuclear energy use

"Basic requirements

for justification of strength

of

VVER type reactor internals" approved by the

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

ABBREVIATIONS

NPP – nuclear power plant

VVER – pressurized water reactor

RI – reactor internals

DI – dynamic impacts

UEZ – ultimate embrittlement zone

AO – abnormal operation of RI

NO – normal operation of RI

SAR - safety analysis report

DBA – design basis accident

RF – reactor facility

Appendix 2

to federal rules and regulations

in the field of nuclear energy use

"Basic requirements

for justification of strength

of

VVER type reactor internals" approved by the

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

TERMS AND DEFINITIONS

The following terms and definitions are used in these Basic requirements.

1. Internals – in-vessel shaft, protective tube unit, reflection shield, basket <2>, shaft bottom <2>, attachments of internals, except for hold-down assemblies.

--------------------------------

<2> For reactor facility of VVER-400 type.

2. Local loss of stability is local buckling of individual components of internals due to compression and/or shear stresses.

3. Abnormal operation of internals is AO of the NPP affecting the state of internals, requiring emergency reactor shutdown and not resulting in an accident.

4. Unstable development of a crack is development of a crack in a metal which does not require an increase in load.

5. Nominal allowable stress is a conventional value equal to the maximum allowable stress in uniaxial stress state, which is determined by the strength limit and yield limit at the design temperature.

6. Normal operation of internals in operation of internals in all modes which do not require emergency reactor shutdown.

7. Global loss in stability is a loss in stability of a component of internals due to longitudinal compression loads and/or torque with bending or torsion of the entire component of internals.

8. General bending stresses are stresses caused by the impact of mechanical loads and varying from the maximum positive value to the minimum negative value throughout the cross-section area.

9. General membrane stresses are stresses caused by the impact of mechanical loads uniformly distributed throughout the cross-section area and equal to the average value of stresses due to these loads in this cross-section.

10. General temperature stresses are stresses caused by nonuniform distribution of temperatures by volume of the component under consideration or by the difference in linear expansion coefficients of materials.

11. Postulated crack is a defect artificially introduced in the design arrangement in the form of a crack (through, elliptical, semi-elliptical or quarter-elliptical) for calculation of unstable development of a crack or the calculation of the kinetics of its growth and determination of its dimensions as of the end of the design operating life.

12. Ultimate embrittlement of materials is achievement of the state of metal in the irradiated zone of a component of internals which may result in its quasi-brittle fracture due to irradiation, in absence of plastic deformations.

13. Limit state is a state of a structural component of internals, exceedance of which during operation will lead to a loss of integrity, displacements exceeding the design values, or the start of metal destruction mechanisms.

14. Reduced stress is an equivalent value of stress reduced to conditions of uniaxial stress state, used for assessment of strength.

15. Design basis accident is an accident affecting the state of internals, with initiating events and end states established in the NPP design and designated safety systems ensuring mitigation of consequences within the established limits in case of any failure of a safety system component independent of an initiating event and considered in the NPP design or in case of a human error independent of an initiating event.

16. Radiation-induced creep is a process of accumulation of irreversible deformation of metal of internals over time in case of joint impact of irradiation and load.

17. Radiation-induced swelling is an increase in volume of a material due to neutron irradiation.

13. Design temperature is the maximum average integral temperature value by the wall thickness (cross-section) of a structural component of internals in the loading mode under consideration.

14. Structural element of internals is a part of RI, where a separate calculation model is used for calculation of strength.

Appendix 3

to federal rules and regulations

in the field of nuclear energy use

"Main requirements

for justification of strength

of

VVER type reactor internals" approved by the

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

CONVENTIONS

Appendix 4

to federal rules and regulations

in the field of nuclear energy use

"Main requirements

for justification of strength

of

VVER type reactor internals" approved by the

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

CLASSIFICATION

OF STRESSES IN COMPONENTS OF INTERNALS

1. Classification of stresses during a calculation of static strength of components of internals and DI shall be performed for the following stress categories:

general membrane stresses;

local membrane stresses;

general bending stresses;

local bending stresses;

general temperature stresses;

local temperature stresses;

local bearing stresses;

local shear stresses.

2. Various categories of stresses are grouped (groups of stress categories) depending on the type, nature of loads and calculation purposes.

3. During a calculation, reduced stresses in each group shall be determined and compared with the corresponding allowable stresses.

4. Based on the analysis of stresses due to mechanical loads and temperature impacts, the most stressed areas of components of internals, as well as areas with greatest exposure and radiation-induces swelling shall be selected for strength assessment.

5. Design groups of stress categories ()₁ and ()₂ used for calculations of static strength and DI are presented in Table 1 of this Appendix.

Table 1

Design groups of stress categories

6. Examples of grouping of stress categories ()₁ and ()₂ in various components of internals are presented in Table 2 of this Appendix.

Table 2

Examples of grouping of stress categories()₁ and ()₂

in components of internals

Appendix 5

to federal rules and regulations

in the field of nuclear energy use

"Main requirements

for justification of strength

of

VVER type reactor internals" approved by the

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated October 5, 2017 No. 409

Table 1

Combinations of design loads and

allowable stress values for structural components of internals in case of external DI

Table 2

Load combinations and allowable stresses for fasteners

of internals in case of external DI