Переводы документов. Translations in English

NP-009-17. Rules of nuclear safety of research reactors

Approved by

Approved by

Order of the Federal

Environmental, Industrial

and Nuclear Supervision Service

dated August, 4, 2017 No. 295

 

FEDERAL RULES AND REGULATIONS

IN THE FIELD OF NUCLEAR ENERGY USE

"RULES OF NUCLEAR SAFETY OF RESEARCH REACTORS"

(NP-009-17)

 

I. Purpose and scope

 

1. These Federal regulations and rules in the area of nuclear energy use "Rules of nuclear safety of research reactors" (NP-009-17) (hereinafter - the Rules) have been developed in accordance with Article 6 of Federal Law No. 170-FZ "On nuclear energy use" dated November, 21, 1995, Decree of the Government of the Russian Federation No. 1511 dated December, 1, 1997 "On approval of the Regulation on development and approval of Federal rules and regulations in the area of nuclear energy use" (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 shall be applicable to designed, constructed and operated nuclear research reactors (hereinafter - research reactors) except for pulse nuclear research reactors.

3. These Rules establish the requirements for the design, characteristics and operating conditions for systems and components of research reactors as well as the organizational requirements aimed to ensure nuclear safety of research reactors.

4. These Rules are developed on the basis of safety assurance principles and requirements established in the federal regulations and rules "General safety provisions of nuclear research installations" (NP-033-11) approved by Order of the Federal Environmental, Industrial and Nuclear Supervision Service No. 348 dated June, 30, 2011 (registered by Ministry of Justice of the Russian Federation on August, 29, 2011; registration number No. 21700; Russian Newspaper, 2011, No.195).

5. Abbreviations and designations used in these Rules are given in Appendix 1, and terms and definitions are given in Appendix 2 to these Rules.

 

II. General provisions

 

6. The aim of RR nuclear safety assurance is to prevent appearance of any conditions for occurrence of a nuclear accident, to prevent unauthorized transition of the RR to the critical state and the RR power increase above the safe operation limits established in the design and engineering documentation (hereinafter - the design) of the research reactor, to eliminate any SSCR in managing nuclear materials and to prevent damages of the components containing nuclear materials.

7. The RR nuclear safety shall be ensured by the system of technical solutions and administrative measures particularly:

compliance of the engineering and technical solutions used in the RR project with the requirements of federal regulations and rules in the area of nuclear energy use and state of the art in science, technology and production;

use of the RR inherent self-protection properties;

adherence to the RR project requirements in the course of the RR construction and development;

usage of RR passive safety systems and components arranged on the basis of independence, diversity, redundancy and single failure principles;

application of practice-proven technical solutions and justified methods, usage of calculation and experimental studies of the RR neutron and physical and thermophysical  characteristics;

implementation of the quality assurance systems, compliance standards, qualification of the personnel, formation and implementation of safety culture at all stages of the RR life cycle.

 

III. Requirements for the research reactor project

aimed to ensure nuclear safety

 

General requirements

 

8. Safety important RR systems and components shall be designed with due regard for any mechanical, thermal, chemical, radiation and other internal impacts possible under normal and abnormal operation including design basis accidents as well as external impacts of natural and human-induced origin that can occur at the RR deployment site.

9. When designing safety important RR systems (components), preference shall be given to the systems (components) designed on the basis of the passive operation principle and inherent self-protection properties and also on implementation of safe failure and single failure principles.

10. The following information shall be specified in the RR project and operation documentation:

neutron and physical, thermohydraulic and any other characteristics affecting nuclear safety;

conditions and frequency of the NRF neutron and physical characteristics checks for compliance with the project;

reliability parameters of the RR normal operation systems important for the NRF nuclear safety and their components referred to safety classes 1, 2 and 3, as well as safety systems and their components;

conditions, scope and frequency of maintenance (administrative and technical works aimed to maintain the structures, systems and components in the state complying with the requirements of the design, federal regulations and rules in the area of nuclear energy use) and inspection of safety important systems;

information on validation and verification of software tools used to substantiated nuclear safety of the research reactor;

control and testing programs and procedures in the course of installation, adjustment, operation and decommissioning of safety important systems (components);

technical and administrative measures aimed to ensure nuclear safety of the research reactor in the course of testing, replacement and withdrawal of the CPS CD actuators and other reactivity control devices from service for repair;

procedures for determination of reactivity margin, efficiency of the CPS control devices and sub-criticality of the reactor;

the procedure for determination of the reactor thermal power;

the procedure for calibration of the neutron flux density control channels in accordance with thermal power of the reactor;

technical and administrative measures for nuclear safety assurance in the course of managing fresh and spent nuclear fuel;

the lists of controlled and regulated parameters;

the lists and values of the parameters used to generate signals for actuation of the protective safety systems;

the lists of the RR equipment interlocks and protections and the requirements for their actuation conditions;

operation limits and conditions as well as safe operation limits and conditions with due regard for all controlled neutron and physical, thermohydraulic and other characteristics affecting nuclear safety and determined with due regard for the measurement errors, uncertainty of calculations and response time of the control systems;

intensity levels of external impacts requiring actuation of the protective safety systems;

results of analysis of safety important systems responses to internal and external impacts of natural and human-induced origin with due regard for their possible combinations and accompanying with any other interdependent processes;

reactivity margin of the research reactor with assessment of error for the applied calculation methods and with due regard for any possible deviations of the core component parameters from the nominal values in the course of their manufacturing;

aggregate efficiency of the CPS control devices, efficiency of individual CPS control devices and their groups, efficiency of experimental devices and nuclear fuel in the course of the nuclear core refueling;

reactivity effects and coefficients including temperature and power reactivity effects and also barometric and density reactivity effects (where necessary) and reactivity effects due to fuel burn-up and the reactor poisoning;

the list of nuclear-hazardous works in the course of the RR operation and nuclear safety measures in the course of these works including refueling works;

the list of initiating events for design basis accidents and the list of beyond design basis accidents (particularly due to complete blackout of the research reactor, loss of the ultimate heat sink, aircraft crash) and also analysis results for design basis and beyond design basis accidents and their consequences;

the list of parameters, procedures and criteria used for assessment of the residual lifetime and replacement of safety important components;

design (specified) service life of safety important components and design (specified) service life of the RR.

11. The equipment provided in the RR project to ensure the RR nuclear safety shall include:

reactivity control devices used to control the reactor power including reactivity compensation control devices, manual controller devices and automatic controller devices in case of automatic power control;

the normal operation control system ensuring variation of the reactor power level up to the preset one, maintenance of the preset power level and scheduled shutdown of the research reactor;

protective safety systems including the set of safety systems performing the emergency protection function for the research reactor (hereinafter - the emergency protection) and the emergency core cooling system;

the control safety system providing control of safety systems;

emergency warning signal output devices (the alarm horn with a distinguishable audio tone) in any cases prescribed in the RR project;

alarm signal output devices (light and acoustic) upon reaching of the parameter setpoints and conditions for the EP actuation;

warning signal output devices (light and acoustic) in case of any abnormal operation of the RR systems and components;

output devices for indication signals of voltage presence in the power supply circuits and the state of equipment and instruments;

back-up power supply sources for systems and components used during the RR scheduled shutdown and subsequent cooldown in case of any failures of the main (operating) power supply sources;

emergency power supply sources ensuring operation of at least two power level control channels and position indicators of the CPS control devices, monitoring of the temperature regime in the reactor core and the SNF storage, emergency cooling of the core and functioning of the emergency control room.

The possibility of the RR operation without the automatic power control system shall be substantiated in the RR project.

12. Technical solutions used in the RR project shall provide:

sub-criticality of the reactor after lifting of the EP control devices - at least 1% (Кeff base_1_256205_32768 0.99) at any reactor campaign moment;

sub-criticality of the reactor with all CPS control devices inserted - at least 2% (Кeff base_1_256205_32769 0.98) at any reactor campaign moment;

sub-criticality of the reactor in the long-term shutdown mode (with all CPS control devices inserted and with the use of any other reactivity control devices or partial unloading of nuclear fuel from the nuclear core) - at least 5% (Кeff base_1_256205_32770 0.95);

state monitoring for the reactor and safety important systems particularly in case of design basis accidents;

protection of any software and hardware used in the control systems against unauthorized access;

integrity and operability of the equipment used for recording and storage of information necessary for identification of the initiating events of design basis accidents and determination of operation algorithms for safety important systems and actions of the personnel under the design basis accident conditions.

 

Nuclear core and normal operation

safety important systems

 

Nuclear core

 

13. Design of the reactor under normal and abnormal operation including design basis accidents shall prevent any unauthorized changes of the nuclear core composition and configuration resulting in reactivity increase and (or) deterioration of heat removal and subsequent damage of fuel elements in excess of the design limits.

14. Design of fuel assemblies and fuel elements, nuclear fuel, materials of the fuel element cladding shall ensure non-exceedance of the design limits for fuel element damages under normal and abnormal operation including design basis accidents with due regard for:

physical and chemical interaction between the fuel element claddings and nuclear fuel, fuel element (fuel assembly) claddings and the coolant;

shock and vibration impacts, thermal cycle loading, fatigue and material ageing;

impact of fission products and impurities in the coolant on strength and corrosion resistance of the fuel element and fuel assembly claddings;

radiation impact and any other factors deteriorating mechanical characteristics of the fuel element and fuel assembly materials.

15. Design and configuration of the nuclear core shall eliminate positive power and temperature reactivity coefficients within the entire change range of the reactor parameters under normal and abnormal operation including design basis accidents.

16. Design of the nuclear core or the reflector shall ensure the possibility to install an external (start-up) neutron source used for physical start-up and subsequent operation of the research reactor (if this is contained in the project).

17. The RR project shall include thermal and technical reliability analysis for the nuclear core in order to substantiate sufficiency of the provided margins to the safe operation limits.

18. The nuclear core shall be designed in such a way so that to prevent any jamming and spontaneous movement of the CPS control devices causing positive reactivity (CPS CD ejection).

19. Any hardware and methods for leak-tightness monitoring of fuel elements (fuel assemblies) in the shut-down and power-operated reactor provided for in the RR project shall ensure reliable and timely detection of any leaky fuel elements (fuel assemblies).

20. The RR project shall define correspondence between damages of fuel elements and activity of the primary circuit coolant with regard to reference radionuclides (with due regard for efficiency of the system for the coolant purification from fission products).

21. Fuel elements (fuel assemblies) with different nuclear fuel (particularly with different enrichment), special burnable neutron poisons, fuel elements with burnable neutron poison, fuel elements with combined nuclear fuel shall have labeling (distinguishing signs) which shall be preserved within their entire service life and subsequent storage.

 

Experimental devices

 

22. Design of experimental devices shall preclude any possibility for spontaneous movement of any components of experimental devices and irradiated items in the course of their operation within the reactor and also provide confinement of radioactive substances in case of any breakage of these experimental devices and irradiated items through the use of the RR safety barriers including localizing systems.

23. Loading (unloading) of any replaceable components of experimental devices and reactor testing objects accompanied with positive reactivity insertion exceeding 0.3base_1_256205_32771eff shall be performed with the reactor shut down.

24. In case it is necessary to load (unload) any reactor testing objects in the course of the reactor power operation the necessity for works under these conditions shall be substantiated in the design, and it should be preliminarily confirmed through experiments that the inserted positive reactivity in the course of loading (unloading) does not exceed 0.3base_1_256205_32772eff.

25. The positive reactivity insertion rate in the course of installation (unloading) of any reactor testing objects with the efficiency of more than  0.3 base_1_256205_32773eff shall not exceed 0.07 base_1_256205_32774eff/s.

26. In case installation (unloading) of any reactor testing objects results in reactivity increase by 0.7 base_1_256205_32775eff and more step-by-step reactivity increase with the interval not exceeding 0.3 base_1_256205_32776eff shall be provided.

27. It is permitted to perform operations for installation (unloading) of the reactor testing objects without any limitations for the positive reactivity insertion interval and rate in case Кeff does not exceed 0.95 prior to the works, in the course of the works and after their completion.

28. Conditions, scope and frequency of inspections of experimental devices for compliance with the design characteristics shall be defined in the RR project.

29. Results of calculation estimates and experimental assessment of the impact of experimental devices on reactivity, energy emission distribution in the nuclear core and efficiency of the CPS control devices shall be presented in the RR project and the SAR.

30. Nuclear safety analysis for application of any new or modified experimental devices requiring introduction of changes to the RR SAR shall be agreed by the operating organization with the main engineering organization. In case there is no necessity to introduce any changes to the RR SAR the operating organization shall inform the main engineering organization on the use of any new experimental devices or modification of the existing ones.

 

Nuclear core cooling system under normal

operation conditions (the primary circuit)

 

31. The normal operation nuclear core cooling system (the primary circuit) shall ensure heat removal from the nuclear core without any deviations from the established design limits with regard to temperature and temperature change rate for the core components and also for experimental devices.

32. The following shall be provided in the RR project:

results of strength calculations for the reactor vessel and the primary circuit pipelines;

the primary circuit reliability analysis results with due regard for any internal impacts possible under normal and abnormal operation including design basis accidents as well as external impacts of natural and human-induced origin;

permissible movements and vibrations of the primary circuit pipelines and structural elements under normal RR operation conditions.

33. Technical solutions used in the RR project shall take into account the possibility for deterioration of heat transfer characteristics of the heat exchanging equipment in the course of operation.

34. The RR project shall provide for the following:

protection against impermissible increase or reduction of temperature, pressure or coolant flow in the primary circuit in case of abnormal operation;

compensation of any coolant volume variations in case of any changes of the coolant specific density;

means for detection of any coolant losses in case of leakages, means for compensation of the coolant losses through leakages and means for the primary circuit protection against unauthorized drainage of the coolant;

coolant purification from any impurities including radioactive ones;

monitoring of the core cooling system parameters.

35. Chemistry of the coolant and the permissible radionuclide content in the course of operation as well as the coolant quality parameters shall be defined in the RR project.

36. Any technical solutions applied in the research reactor and its physical properties shall prevent exit of the shut-down reactor from sub-critical state upon actuation (switch-off) of the primary circulation pumps.

 

Normal operation control system

 

37. The normal operation control system shall include:

controls for the actuators of the CPS control devices, loading and experimental devices;

position (state) monitoring means for the MC CDs, AC CDs and RC CDs;

position (state) monitoring means for irradiated items with actuators for remote movement;

at least two independent channels for monitoring of the neutron flux density with indicating devices, in this case the possibility for recording of the time-based neutron flux density changes shall be provided;

at least two independent neutron flux density change rate (period) control channels with indicating devices;

channels for monitoring of the parameters for safety important RR process systems.

38. Change ranges and rates for the regulated process parameters under normal and abnormal operation including design basis accidents shall be substantiated in the RR project.

39. The neutron flux density control range of the normal operation control system shall cover the entire reactor power variation range defined in the RR project, and in case neutron flux density and/ or density change rate control channels are operated within limited sub-ranges of neutron flux density measurement the two adjacent controlled sub-ranges shall overlap by at least a decade.

40. The reactor and the main RR systems shall be controlled from the RR control room equipped with two-way loudspeaker communication with the reactor hall and other RR rooms. The RR control room shall be equipped with telephone and radio communications.

41. In case the control channels specified above in par. 37 do not ensure neutron flux density control in the course of nuclear fuel loading the reactor shall be equipped with the additional (start-up) neutron flux density control system. This system may be removable and installed for the period of nuclear fuel loading and shall include at least two neutron flux density controlled channels with indicating and recording devices. Characteristics and parameters of the additional neutron flux density control system shall be specified and substantiated in the design and in the SAR.

42. The normal operation control system shall prevent:

insertion of positive reactivity by any reactivity control devices in case the EP control devices are not lifted;

insertion of positive reactivity at the rate exceeding 0.07base_1_256205_32777eff/s;

insertion of positive reactivity by any reactivity control devices in the presence of any warning alarms with regard to the neutron flux density or the neutron flux density increase rate (period) or the control channels for the parameters of safety important RR process systems;

insertion of positive reactivity by any reactivity control devices in case of power supply loss in the intermediate position indicator circuits of the control device used for reactivity increase or in the emergency and warning alarm circuits;

the possibility for remote reactivity increase from two and more workplaces and (or) by two or more ways at the same time.

43. The normal operation control system shall ensure:

step-by-step insertion of positive reactivity with the interval not exceeding 0.3 base_1_256205_32778eff for any reactivity control devices with the efficiency of more than 0.7 base_1_256205_32779eff, in this case a temporary pause with the duration specified in the RR project (operation documentation) shall be made after each reactivity insertion step;

insertion of negative reactivity upon the EP signal at the maximum possible rate through the use of RC CDs, MC CDs, AC CDs and any other reactivity control devices;

the possibility to break the power supply circuit of the drives of the CPS control devices with the efficiency exceeding 0.7 base_1_256205_32780eff from the RR control room, in this case the motor power supply circuit break shall not affect the possibility to bring the RR to sub-critical state upon the EP signal;

automatic termination of positive reactivity insertion by the loading and experimental devices upon the EP signal, and in case of necessity - automatic reduction of reactivity caused by the loading or experimental devices;

warning alarm with regard to the neutron flux density and the neutron flux density increase period with the setpoint of at least 20 s;

generation of signals for the control safety system in case of any malfunctions of this system components in accordance with the design;

operability check including veriffication of the light and acoustic alarm systems.

44. The reactor power range within which control is arranged through the use of the automatic controller shall be defined and substantiated in the RR project; characteristics of the automatic power control system and estimated error in maintenance of the required power level shall be specified, and absence of any power self-oscillations shall be confirmed.

45. If several neutron flux density control channels are used in the automatic control system any changes of the reactor power by the automatic control system in case of any neutron flux density control channel disabling or failure shall be prevented.

 

Safety systems

 

Emergency protection and other shutdown systems

 

46. The reactivity control devices used in case of any RR abnormal operation shall include at least two independent EP control devices or groups of EP control devices depending on the RR project.

47. Efficiency of the EP control devices without the most effective EP control device (group of EP control devices) and their response time shall ensure:

the reactor power reduction rate sufficient to prevent damages of fuel elements and experimental devices in excess of the safe operation limits;

transition of the reactor into the sub-critical state and its maintenance in this state within the time period sufficient for insertion (actuation) of the slower CPS control devices.

48. The EP control devices shall have end position indicators.

49. In case of an emergency signal the EP control devices shall be actuated automatically from any position. Negative reactivity insertion shall be ensured at any section of the EP CD movement; in this case negative reactivity shall be also inserted by other CPS control devices.

50. The EP CD actuator shall perform its safety functions regardless of the state of its power supply sources.

51. Apart from emergency shutdown of the RR the EP control devices may be also used for scheduled RR shutdown in case of necessity.

52. Apart from EP control devices the RR project may provide for any other automatically or remotely actuated shutdown systems among the protective safety systems.

53. Shutdown systems shall ensure maintenance of the reactor in sub-critical state with due regard for potential reactivity release particularly due to temperature and power reactivity effects.

54. Reliability analysis and parameters for the CPS shall be provided in the design and the RR SAR. Reliability analysis shall be performed with due regard for common cause failures and human errors.

 

Emergency nuclear core cooling system

 

55. The safety system ensuring emergency cooling of the nuclear core in case of any failure of the forced cooling system shall be provided in the RR project for the reactor with the forced core cooling system.

56. Technical solutions applied in the RR project shall ensure the maximum use of natural coolant circulation in the course of emergency RR cooling possible for this RR project.

57. The following shall be substantiated in the RR project: the list of parameters for actuation of the emergency core cooling system, setpoint values and conditions for the system activation for all initiating events of design basis accidents.

58. Actuation and disabling of the emergency core cooling system shall not result in the RR exit from sub-critical state.

59. The possibility to control the emergency cooling process both from the main RR control room and from the emergency control room shall be provided.

 

Control safety system

 

60. The control safety system shall include at least two independent protection channels for the neutron flux density and two independent protection channels for the neutron flux density increase rate (period).

61. Sensitivity and location of the neutron flux detectors of the control safety system shall be selected in such a way so that to ensure the possibility for the EP actuation in the course of the RR transition to critical state and at any power level within the range defined in the RR project.

62. In case protection channels operated in the limited sub-range of neutron flux density are used overlapping of the sub-ranges within at least one decade shall be provided.

Switching of the sub-ranges shall be automatic and shall not impair generation of EP signals.

63. In case of any structural, electrical or functional combination (consolidation) of the measuring sections of the control safety system protection channels with each other and/ or with the measuring sections of the normal operation control system channels it should be demonstrated in the RR project that such combination would not affect the EP capability to perform safety functions.

64. Protective function for each process parameter for which the EP actuation or switching to emergency cooling of the nuclear core is prescribed in the design within the entire reactor parameter change range shall be implemented by at least two independent channels.

65. Upon lifting of EP control devices with the efficiency exceeding 0.7 base_1_256205_32781eff step-be-step insertion of positive reactivity (step motion) with the interval of not more than 0.3 base_1_256205_32782eff shall be provided; in this case the positive reactivity insertion rate shall not exceed 0.07 base_1_256205_32783eff/s.

66. The control safety system shall prevent the EP CD lifting in the following cases:

MC CDs, AC CDs and RC CDs are not at the limit switches corresponding to the maximum RR sub-ctiticality;

there are any emergency or warning alarms with regard to the parameters of safety important RR process systems listed in the RR project.

67. Emergency protection shall be actuated in the following cases:

reaching of the EP actuation setpoint in any neutron flux density protection channel or neutron flux density change rate (period) protection channel;

failure of any neutron flux density protection channel or neutron flux density change rate (period) protection channel in case the number of these channels is not more than two;

reaching of the EP actuation setpoints for safety important process parameters;

any emergency signals from experimental devices requiring the RR shutdown;

initiation of the EP actuation by the personnel through the use of the relevant buttons (keys);

failure of power supply for control safety systems, particularly for the power supply units of the neutron flux density detectors in the control or protection channels.

68. In case the number of EP channels for any parameter is more than two the EP actuation subject to simultaneous signals from any two EP channels for this parameter is permitted.

69. The control safety system shall provide for output of light and acoustic emergency alarm signals in the RR control room in order to inform the operator on inoperable state of any protection channels, reaching of the emergency setpoints and the EP actuation.

The control safety system shall generate and transmit to the emergency RR control room any signals necessary to ensure performance of its functions.

70. The EP actuation setpoints and conditions shall prevent reaching of the safe operation limits, in this case the emergency setpoint for the neutron flux density increase period shall be at least 10 s.

71. Continuous diagnostics of the good operable state of the protection channels with display of information on any failures in the RR control room shall be provided, and in case of any channel failure or withdrawal from service an emergency signal from this channel shall be generated automatically.

72. The RR project shall provide for the possibility of the RR shutdown, actuation of the protective safety systems and monitoring of the reactor parameters from the emergency control room in case it is impossible to carry out these activities from the main RR control rooms.

 

IV. Requirements for nuclear safety assurance

in the course of the reactor plant commissioning

 

General requirements

 

73. Compliance with the requirements of the reactor facility project and nuclear safety assurance for the research reactor shall be confirmed in the course of the reactor facility commissioning (the type of activities including commissioning works, physical start-up of the RR and power start-up of the reactor facility when compliance of individual RR systems and equipment with the design is verified).

74. Preparation for the reactor facility commissioning shall be arranged in accordance with the reactor facility commissioning program agreed with the reactor facility project developers and approved by the operating organization.

75. The reactor facility commissioning program shall be developed on the basis of the design and engineering, process, operation and organizational and administrative documentation of the reactor facility.

76. The operating organization shall be responsible for development and implementation of the reactor facility commissioning program.

77. The requirements of the reactor facility commissioning program shall be mandatory for all organizations and entities participating in the reactor facility preparation for commissioning.

78. The following shall be defined in the reactor facility commissioning program:

arrangement of the reactor facility commissioning works, participants of the works, their tasks and responsibilities;

initial state of the reactor facility before commencement of the works in accordance with the reactor facility commissioning program;

stages and sub-stages of the works, scope and requirements for the documentation necessary for their implementation;

initial state of the reactor facility before commencement of the future stage of the reactor facility commissioning works;

qualification requirements for the personnel and their training;

technical and administrative measures aimed to ensure safety at each stage of the works;

the list of reactor facility systems used for physical start-up of the research reactor.

79. The reactor facility commissioning program shall include three successive stages of works:

commissioning works;

physical start-up of the research reactor;

power start-up of the reactor facility.

In accordance with the decision of the operating organization all three stages may be presented within a single program or as three independent documents with due regard for the state of construction and installation works at the reactor facility site and the expected time limits for their completion.

 

Commissioning works

 

80. In the course of commissioning works at the commissioned reactor facility, compliance of the characteristics and parameters of safety important systems (components) with the nominal data and values specified in the design and operability of each safety important system shall be verified, and integrated operability check shall be performed for the systems (components) with due regard for their mutual impacts.

81. Commissioning works for the systems not used for physical start-up of the research reactor may be performed after physical start-up within the preparation works for the reactor facility power start-up subject to the relevant substantiation.

82. Subsequent to the results of testing in the course of commissioning works it should be demonstrated that characteristics and parameters of nuclear safety important systems (components) comply with the design values, in this case:

the CPS channels (additional (start-up) CPS channels) ensure the neutron flux density control in the absence of nuclear fuel and in presence of the start-up neutron source in the reactor core;

EP and other reactor shutdown systems are actuated by simulation of exceedance of the relevant power and period setpoints, setpoints of the process parameters or in case of any malfunctions of the relevant control channels;

response time of the emergency protection and other shutdown systems complies with the requirements specified in the design;

acoustic and light alarms provide the personnel with the relevant information in case of reactor abnormal operation and also in case of any failures of safety important systems.

 

Physical start-up of the research reactor

 

83. Subsequent to acceptance of the reactor facility structures, systems and equipment used for physical start-up by the operating organization the RR readiness for physical start-up shall be checked by the nuclear safety commission appointed by the order of the operating organization and the commission of the Federal Environmental, Industrial and Nuclear Supervision Service.

84. The nuclear safety commission shall check:

compliance of the performed works with the RR project;

results of the commissioning works and testing of the RR systems, availability of the commissioning work completion certificates;

performance of the planned administrative and technical arrangements for the RR nuclear safety assurance;

availability and contents of the operation documentation within the scope required for physical start-up of the research reactor;

preparedness of the personnel for the works according to the RR physical start-up program particularly availability of any permits for performance of works in the area of nuclear energy use and results of the personnel workplace knowledge assessment.

85. Subsequent to verification of the RR readiness for physical start-up by the commission of the Federal Environmental, Industrial and Nuclear Supervision Service and after elimination of all drawbacks revealed by the above-mentioned commission the operating organization shall issue the order on the RR physical start-up.

86. The RR physical start-up shall be performed in accordance with the RR physical start-up program agreed with the reactor facility project designers and approved by the operating organization.

The list and results of the performed works shall be registered in the in-process shift log sheet.

87. The following shall be defined in the RR physical start-up program:

the list of systems and equipment required for the RR physical start-up;

arrangement of the physical start-up works with indication of all participants of the works, their tasks, rights and responsibilities;

initial state of the systems and equipment prior to commencement of the reactor fueling works;

the procedure for the reactor fueling;

the procedure for the critical state reaching;

description of the experiments in order to define the RR characteristics and the procedure for their performance;

expected critical loading of the nuclear core, critical positions (states) of the reactivity control devices, their efficiency, assessment of impact of the loaded fuel, moderating materials and the coolant on reactivity;

the list of methods used to perform experiments and measurements in the course of physical start-up;

nuclear safety measures during physical start-up.

88. Nuclear-hazardous works in the course of the RR physical start-up shall be performed in accordance with the nuclear safety guidelines for the RR physical start-up approved by the chairman of the operating organization and specifying the following:

the procedure for performance of nuclear-hazardous works particularly the procedure for nuclear fuel loading into the core and the procedure for the reactor rising to the critical state;

nuclear safety measures;

design values of critical loadings and efficiencies of the CPS control devices;

estimated impact of the experimental devices and the coolant on reactivity;

safe operation limits and conditions within the period of the RR physical start-up.

89. Physical start-up shall begin with installation of the external (start-up) neutron source to the reactor in accordance with the RR project.

Availability of the reactivity control devices necessary to restore sub-criticality in case of any approach to criticality not provided in the RR physical start-up program shall be ensured in the course of physical start-up.

Neutron flux density and neutron flux density increase rate protection setpoints providing acoustic and light alarms at the minimum power level established in the RR project shall be set for the instruments of the control safety system.

90. Loading of nuclear fuel and/or filling of the moderator after the nuclear fuel loading into the reactor core shall be accompanied with development of count-down curves according to the readings of at least two power control channels. In this case at least two count-down curves shall have safe run.

91. Upon reaching of Keff equal to 0.98 (neutron multiplication factor of 50) sequential efficiency assessment for all CPS control devices shall be performed and presence (absence) of critical state after withdrawal of all CPS control devices shall be checked. Further loading shall be performed in portions not resulting in reactivity increase by more than 0.3 base_1_256205_32784eff, in this case the positive reactivity insertion rate in the course of loading shall not exceed 0.07 base_1_256205_32785eff/s.

In case it is impossible to perform loading of  the portion resulting in reactivity increase by less than 0.3 base_1_256205_32786eff with due regard for the reactor design the portion efficiency shall be minimum permissible for the reactor design, and compliance with the requirements of par. 137 of these Rules shall be ensured.

92. In case the possibility for early (prior to completion of the loading) reactor critical state reaching appears from the results of sub-criticality assessment and/or readings of the neutron flux density control channels the core fueling shall be stopped. Any further works shall be performed in accordance with the written instruction from the physical start-up supervisor approved by the RR (reactor facility) chief engineer (head).

93. The certificate signed by the physical start-up supervisor containing the basic results of physical start-up and information on compliance of the list of performed works with the physical start-up program shall be issued subsequent to the physical start-up.

94. Subsequent to the physical start-up results the report containing the physical start-up results and recommendations for the RR operation documentation updating shall be issued. The report on the physical start-up results shall be approved by the chair of the operating organization.

95. In case the results of physical start-up demonstrate impossibility to reach the design characteristics of the RR and necessity to make any changes in the RR project the operating organization shall introduce the relevant modifications to the design and documents substantiating safe operation of the RR.

 

Power start-up of the reactor plant

 

96. Readiness of the reactor plant for power start-up shall be verified by the commission appointed by the order of the operating organization.

97. Subsequent to elimination of any defects revealed by the commission for verification of the reactor facility readiness for power start-up the operating organization shall issue the order on power start-up of the reactor facility.

98. The power start-up shall be performed in accordance with the power start-up program adjusted in accordance with the physical start-up results, agreed with the reactor facility project designers and approved by the operating organization.

87. The following shall be defined in the power start-up program:

arrangement of the power start-up works, participants of the works, their tasks, rights and responsibilities;

initial state of the reactor facility and any auxiliary systems supporting power start-up before commencement of the works in accordance with the reactor facility power start-up program;

stages and sub-stages of works;

initial state of the reactor facility and safety important systems prior to commencement of each stage of the works;

scope and requirements for the documentation necessary for power start-up;

the list, methods and sequence of the planned experiments and tests at the prescribed power levels;

the procedure and method for the radiation situation measurement;

expected results of experiments and tests;

nuclear and radiation safety measures;

requirements for the personnel;

requirements for safety assurance at each stage of the works;

requirements for preparation of the report on the power start-up results.

100. Upon completion of the power start-up the operating organization shall issue the report containing the following:

results of the studies prescribed in the power start-up program;

recommendations for adjustment of the design, SAR and operation documentation;

basic parameters and characteristics of the reactor included into the RR certificate.

The report on the power start-up results shall be approved by the chairman of the operating organization.

101. The operating organization shall issue the RR certificate on the basis of the reactor facility project and the reports on physical and power start-up. The RR certificate shall be issued in accordance with Appendix 3 to these Rules.

102. In case the results of power start-up demonstrate impossibility to reach the design characteristics of the reactor facility and necessity to make any changes in the reactor facility project the operating organization shall introduce the relevant modifications to the design and documents substantiating safe operation of the reactor facility.

 

V. Requirements for nuclear safety assurance

in the course of the research reactor operation

 

General provisions

 

103. Rights and obligations of officers and structural units of the operating organization, rights and liabilities of the personnel for the RR nuclear safety assurance shall be defined In accordance with the procedure established in the operating organization.

104. The operating organization shall approve the list of documentation effective at the reactor facility with due regard for Appendix 4 to these Rules and ensure availability of this documentation at the reactor facility.

.105. The operating organization shall ensure timely introduction of any amendments to the reactor facility documentation particularly introduction of changes to the SAR, process regulations and any other reactor facility operation documents subsequent to the results of physical and power start-up.

106. Any modifications of the RR project shall be approved by the main scientific organization, the main engineering organization and the main design organization within the relevant scope.

107. The RR operation shall be performed in accordance with the reactor facility process regulations and the RR operation manual (guideline) and also with due regard for the requirements of the operation manuals for the RR systems and components and the guidelines on nuclear safety assurance in the course of storage, handling and transportation of fresh and spent nuclear fuel. The above-mentioned documents shall be updated with due regard for the obtained RR operation experience, implementation of new federal regulations and rules in the area of nuclear energy use, introduction of any changes into the RR process systems and equipment and reviewed at least once per five years.

108. Nuclear-hazardous works on the research reactor including the refueling works shall be performed in accordance with special programs (technical resolutions) approved in accordance with the procedure established in the operating organization.

109. The program (technical resolution) for performance of nuclear-hazardous works shall contain:

the aim of the nuclear-hazardous work, sequence and technique for its performance;

administrative and technical nuclear safety measures in the course of nuclear-hazardous works on the research reactor;

results of calculation estimates or experimental assessment of the impact of planned works on the RR sub-criticality;

the technique for performance and control of correctness for nuclear-hazardous works.

110. The technique for performance of recurrent nuclear-hazardous works on the research reactor when the experimentally determined reactivity changes in the course of these works are known may be included into the RR operation documentation.

111. The neutron flux density and the neutron flux density increase rate shall be controlled in the course of nuclear-hazardous works on the research reactor, and the minimum possible EP actuation setpoints for power level and the maximum possible setpoints for power increase period providing acoustic and light alarms in case of any unauthorized RR power increase shall be set for the instruments of the control safety system.

112. Sufficiency of the administrative and technical arrangements for the RR nuclear safety assurance shall be substantiated in the design and presented in the SAR.

113. The operating organization shall check the RR nuclear safety state at least once per a year. Results of the checks shall be reflected in the annual nuclear and radiation safety assessment report.

 

Start-up and power operation mode

 

114. The procedure for the RR operation in the start-up and power operation mode shall be described in the reactor facility process regulations and the RR operation manual (guideline).

115. Experimental studies in the course of the RR operation in the start-up and power operation mode shall be performed on the basis of the RR experimental research program approved in accordance with the procedure established in the operating organization. The list and results of the performed works shall be registered in the in-process shift log sheet.

116. initial state of the shut-down RR and the RR process systems, the required power level and duration of the reactor operation at this power level as well as nuclear safety measures with due regard for the peculiarities of the future experimental research on the RR shall be specified in the RR experimental research program.

117. At any moment of the RR campaign the personnel shall have the information with regard to the core fuel loading pattern, the RR reactivity margin and efficiency of the CPS control devices.

118. Nuclear-hazardous works including the works for maintenance, routine repair, testing and operability checks for safety important systems may be performed in the start-up and power operation mode.

At least 2% sub-criticality shall be ensured and maintained prior to commencement of the works, in the course of the works and after their completion; in this case the neutron flux density and the neutron flux density increase rate shall be controlled, and EP setpoints providing acoustic and light alarms at the minimum power level established in the RR project shall be set for the instruments of the control safety system.

119. In case any new nuclear-hazardous works are performed on the research reactor  the supervising physicist shall be included into the shift staff by the resolution of the RR (reactor facility) chief engineer (head) in order to carry out surveillance over adherence to the nuclear safety arrangements.

120. The start-up and power operation mode shall be finished with insertion of all CPS control devices into the nuclear core (locking of the CPS control devices in the lower limit switches).

121. In case of any accident in the start-up and power operation mode the top-priority actions of the personnel shall be aimed to ensure sub-criticality of the RR and the nuclear core cooling. The action plan in case of any accident in the start-up and power operation mode shall be defined in the relevant guidelines.

 

Temporary shutdown mode

 

112. At least 2% sub-criticality of the reactor shall be ensured and maintained during the RR operation in the temporary shutdown mode as per the moment of the works commencement, in the course of the works and after their completion.

123. All works in the reactor hall subsequent to the RR transition into the temporary shutdown mode including any works for maintenance, routine repair, testing and operability checks for safety important systems shall be performed by the shift and/or repair personnel in accordance with the program recorded in the in-process log sheet and the approved guidelines (regulations), programs and schedules. The list and results of the performed works shall be registered in the in-process shift log sheet.

Any works resulting in increase of the effective neutron multiplication coefficient for the RR shall be performed under the guidance of the shift supervisor by the personnel having the necessary proficiency level and the relevant permits for performance of such works; in this case the neutron flux density and the neutron flux density increase rate shall be controlled.

124. Subsequent to completion of the works for maintenance, repair or replacement of the components of safety important systems operability of the systems (components) and their compliance with the design characteristics shall be checked.

125. The temporary shutdown mode shall be finished with commencement of the EP CD lifting.

 

Long-term shutdown mode

 

126. The RR long-term shutdown mode shall be introduced by the order of the operating organization when the experimental works are completed and the RR operation in the temporary shutdown mode is not feasible.

127. Prior to making the decision on the RR transition to the long-term shutdown mode the operating organization shall develop the arrangements aimed to ensure nuclear safety of the RR in this mode and to prevent operability loss for the safety important system components.

128. Prior to commencement of the RR operation in the long-term shutdown mode at least 5% sub-criticality (Keffbase_1_256205_32787 0.95) of the reactor shall be provided, and the possibility for supplying power to the actuating mechanisms of the CPS CDs, shutdown systems, the RR experimental and loading devices shall be eliminated in case there is any nuclear fuel in the reactor core.

129. The scope and frequency of state control for the RR in the long-term shutdown mode shall be defined in the RR operation manual (guideline).

130. The procedure for preparation of the RR in the long-term shutdown mode for operation in the start-up and power operation mode shall be defined in the program of the RR preparation for operation approved and agreed in accordance with the procedure established in the operating organization.

131. Completion of the long-term shutdown mode and the possibility for the RR operation in the start-up and power operation mode shall be arranged via the order of the operating organization after verification of the RR readiness for operation in the start-up and power operation mode by the nuclear safety commission of the operating organization.

 

Final shutdown mode

 

132. In the RR final shutdown mode the operating organization shall implement administrative and technical arrangements aimed to prepare the RR for decommissioning including nuclear fuel unloading from the reactor core and removal of nuclear fuel and other nuclear materials from the RR site.

133. In case it is impossible to remove all nuclear materials present in the reactor without any special-purpose equipment not provided in the RR project due to spillage of the nuclear fuel solution or scattering of nuclear materials from broken fuel assemblies and/or any design peculiarities of the reactor arrangements aimed to ensure nuclear safety of the shut-down RR shall be developed and implemented, and the terms of reference for the RR decommissioning design development shall include the works for clean-up and removal of the nuclear materials remaining in the RR.

 

Refueling of the nuclear core

 

134. The procedure for the nuclear core refueling shall be defined in the refueling program and (or) the refueling guidelines, refueling schedules and maps approved and agreed in accordance with the procedure established in the operating organization.

135. Heat removal from the handled fuel assemblies without any exceedance of the permissible fuel element temperature parameters established in the RR project shall be provided in the course of refueling.

136. Administrative arrangements and engineering features in the course of refueling works shall prevent any ingress of foreign objects into the equipment, valves and pipelines.

137. The personnel shall have information on any possible RR sub-criticality changes in the course of refueling and upon completion of the planned works as per commencement of each refueling stage based on the performed calculation and experimental assessment.

138. In accordance with the RR project peculiarities the nuclear core refueling shall be performed with the inserted EP control devices and sub-criticality of at least 2% as per the moment of refueling commencement, in the course of refueling and upon its completion or with lifted EP control devices and sub-criticality of at least 1% as per the moment of refueling commencement, in the course of refueling and upon its completion.

139. Engineering features and administrative arrangements shall be provided in the reactors where the required sub-criticality in the course of refueling is ensured by the liquid poison solution in order to prevent supply of coolant without poison to the reactor and the primary circuit in the course of the RR refueling.

140. In case there are no experimental data confirming compliance of the reactor with the requirements of par. 138 of these Rules the refueling program (guideline) shall provide for experimental assessment of the reactor sub-criticality in the course of refueling.

141. Sub-criticality verification in the course of the RR refueling shall be also performed in case the readings of the neutron flux density control devices do not confirm the expected state of the reactor. Subsequent refueling works shall be performed in accordance with the adjusted work program and the written permission of the RR chief engineer (head).

 

VI. Nuclear materials management

 

142. The RR nuclear materials shall be stored in the rooms specified in the RR project and complying with the requirements of federal regulations and rules in the area of nuclear energy use establishing the requirements for safety in the course of NM management at nuclear facilities.

143. All works with nuclear materials at the RR shall be performed by at least two workers.

144. Stationary location of fuel elements, fuel assemblies and containers with nuclear materials shall be arranged for storage of nuclear materials in temporary and permanent storage facilities in order to prevent their unauthorized movement. Кeff value for the storage facility shall not exceed 0.95 under normal and abnormal operation including design basis accidents (particularly with water ingress to the storage facility).

145. Temporary storage facilities for nuclear materials located in the RR rooms shall not affect neutron and physical characteristics of the reactor.

146. Research reactors where any configuration and/or reconfiguration of fuel assemblies as well as loading of nuclear materials to experimental devices and their unloading shall be performed in accordance with the experiment conditions shall be equipped with the proper workplaces for safe performance of these works.

147. The procedure for works with nuclear materials and nuclear safety measures for both NM storage facilities and areas for configuration and (or) reconfiguration of fuel assemblies shall be defined in the nuclear safety guidelines for storage, handling and transportation of fresh and spent nuclear fuel and shall comply with the requirements established in the federal regulations and rules determining safe managing of nuclear materials.

 

 

 

 

 

Appendix 1

to federal rules and regulations

in the field of atomic energy use

"Nuclear Safety Rules for

Research Reactors",

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

dated August, 4, 2017 No. 295

 

LIST OF ABBREVIATIONS AND DESIGNATIONS

 

EP

-

Emergency Protection

AC

-

Automatic Controller

RR

-

Research Reactor

NRF

-

Nuclear Research Facility

RC

-

Reactivity Compensator (compensation device)

Кeff

-

Effective neutron multiplication factor

SAR

-

safety analysis report

SNF

-

Spent Nuclear Fuel

CD

-

Control Device

MC

-

Manual Controller

RF

-

reactor facility

CPS

-

Control and Protection System

SSCR

-

Self-Sustaining Chain Reaction

OO

-

Operating Organization

NM

-

Nuclear Materials

base_1_256205_32788eff.

-

effective fraction of delayed neutrons

 

 

 

 

 

Appendix 2

to federal rules and regulations

in the field of atomic energy use

"Nuclear Safety Rules for

Research Reactors",

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

dated August, 4, 2017 No. 295

 

TERMS AND DEFINITIONS

 

The following terms and definitions are used in these Requirements:

Safe run of the count-down curve - the type of the count-down curve when the extrapolated value of the parameter used for the critical state achievement corresponding to the critical state of a multiplying system  is understated.

Lifting of the control devices of the control and protection system and other reactivity control devices - change in the position (state) of the CPS control devices and other reactivity control devices resulting in positive reactivity insertion.

Loading devices of the research reactor - transport and handling equipment, mechanisms and devices used for loading  of nuclear fuel (refueling) and (or) installation (removal) of experimental devices into the reactor core.

Control channel - a set of sensor(s), communication line and means for signal processing and  information displaying intended to ensure the parameter control.

Independent control channels - control channels without any common (combined) components where a failure of any channel would not result in a failure of another one.

Count-down curve of a multiplying system - dependence of the system reverse multiplication from the parameter changing its multiplying properties.

Reactivity margin of the research reactor - positive reactivity which can be effected in the research reactor with all control devices of the control and protection system and any other reactivity control devices (including remotely moved experimental devices) lifted to the maximum efficiency.

Maximum permissible power - the maximum power not exceeding the rated design power  and permitted for the RR operation due to the existing limitations including the limitations established subsequent to the power start-up results and restrictions in the license conditions.

Shutdown of the research reactor - transfer of the research reactor from the critical (super-critical) state to the sub-critical state through the use of the CPS control devices and other shutdown systems and operation of the research reactor in the sub-critical state.

Emergency shutdown of the research reactor - transfer of the research reactor from the critical (super-critical) state to the sub-critical state through the emergency protection actuation.

Scheduled shutdown of the research reactor - transfer of the reactor from the critical (super-critical) state to the sub-critical state through the use of manual reactivity control devices and reactivity compensation devices with subsequent or simultaneous insertion of the emergency protection control devices.

Nuclear core refueling (reloading) - nuclear-hazardous works on the RR for the nuclear fuel solution filling (drainage), withdrawal and transportation of fuel assemblies, components of experimental devices and tested samples, reactivity control devices and other reactivity-affecting elements for the purpose of their repair, replacement or dismantling.

Commissioning works - the stage of the reactor facility commissioning when operability and compliance with the design are verified for each reactor facility system separately and integrated check of systems in the course of their interaction is performed.

Control device of the control and protection system - a reactivity control device used in the control and protection system and ensuring reactivity variation through its position (state) changing.

In accordance with their functional purpose the CPS control devices are divided into the emergency protection control devices (EP CDs), manual reactivity control devices (MC CDs), automatic reactivity control devices (AC CDs) and reactivity compensation control devices (RC CDs).

Reactivity control devices - the CPS control devices, loading, experimental and other devices changing the RR reactivity in case of their movement or state variation.

Step-by-step movement of reactivity control devices - alternation of the reactor sub-criticality decrease and the subsequent temporary pause sufficient for the power stabilization at the level corresponding to the new sub-criticality level. Every sub-criticality change step shall be initiated by the operator.

Experimental devices of the research reactor - equipment and devices of the research reactor intended to perform experimental research on the reactor including serpentine channels, neutron traps, beam holes, sample holes and also tested items and any devices for their installation.

Nuclear safety of the research reactor - the ability of the RR to prevent nuclear accidents and to limit their consequences.

 

 

 

 

 


 

Appendix 3

to federal rules and regulations

in the field of atomic energy use

"Nuclear Safety Rules for

Research Reactors",

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

dated August, 4, 2017 No. 295

 

(recommended model)

 

                    Research reactor certificate

 

1. RR type and name ____________________________________________________

2. Purpose _____________________________________________________________

3. Location ____________________________________________________________

4. Operating organization ______________________________________________

5. Project developers __________________________________________________

6. Main engineering organization  ______________________________________

7. Main design organization ____________________________________________

8. Main scientific organization ________________________________________

9. RR commissioning date _______________________________________________

10. Specified service life, years ______________________________________

11. Main RR parameters:

rated design power, MW _________________________________________________

maximum permissible power, MW __________________________________________

nuclear core shape and dimensions, mm __________________________________

type of fuel assemblies ________________________________________________

nuclear fuel (nuclide composition, enrichment, %) ______________________

moderator ______________________________________________________________

reflector ______________________________________________________________

coolant ________________________________________________________________

12. Main neutron and physical characteristics of RR:

reactivity margin, base_1_256205_32789     ______________________________________________

                     eff.

prompt neutron lifetime, s _____________________________________________

effective fraction of delayed neutrons _________________________________

power reactivity coefficient, base_1_256205_32790   /MW _________________________________

                               eff.

temperature reactivity coefficient, base_1_256205_32791  /°C ____________________________

                                     eff.

13. Energency cooling systems of the RR ________________________________

14. CPS CD characteristics _____________________________________________

 

CPS CD functional purpose

Number of CD groups, pcs.

Number of CDs in a group, pcs.

Group efficiency change range, base_1_256205_32792eff

Reactivity increase rate upon lifting,

base_1_256205_32793eff/s.

Time of the CPS CD insertion to the core upon EP signal, s

EP

 

 

 

 

 

AC

 

 

 

 

 

MC

 

 

 

 

 

RC

 

 

 

 

 

 

15. Emergency protection with regard to neutron flux density ________________

_____________________________________________________________________________

                  (number of channels and type of instruments)

16. Emergency protection with regard to neutron flux density increase period_

_____________________________________________________________________________

                   (number of channels and type of instruments)

17. Neutron flux density control channels ___________________________________

_____________________________________________________________________________

                 (number of channels and type of instruments)

19. Neutron flux density increase period control channels ___________________

_____________________________________________________________________________

                   (number of channels and type of instruments)

19. Additional reactivity control systems and their efficiency ______________

_____________________________________________________________________________

                 (type, response time, efficiency)

20. Experimental devices and reactivity introduced by them, base_1_256205_32794    ____________

                                                              eff.

_____________________________________________________________________________

21. Additional information __________________________________________________

22. The certificate is issued on the basis of _______________________________

 

"__" __________ ____                                       Chairman

                                               of the operating organization

 

                                                         L.S. (if any)

                                                     _____________(full name)

                                                        signature

 

 

 

 

 

Appendix 4

to federal rules and regulations

in the field of atomic energy use

"Nuclear Safety Rules for

Research Reactors",

approved by the order of the Federal

Environmental,

Industrial and Nuclear Supervision Service

dated August, 4, 2017 No. 295

 

RECOMMENDED LIST

OF THE RESEARCH REACTOR DOCUMENTATION

RELATED TO NUCLEAR SAFETY ASSURANCE

 

1. The list of regulatory documents in the area of nuclear energy use effective at the RR.

2. Technical design and any other technical documentation of the RR including descriptions, certificates, drawings, diagrams and test reports of safety important systems and components.

3. The list of regulations and guidelines effective at the RR with indication of their validity period.

4. The RR safety analysis report.

5. The reactor facility process regulations.

e) crane operation manual (guideline);

7. Operation manual for experimental devices.

8. Operation guidelines for RR systems.

9. Program of experimental research on the RR.

10. Action plan (guideline) for protection of the workers (personnel) in case of  any accident at the RR.

11. Job descriptions for the RR personnel.

12. Orders (extracts from orders) on appointment of the RR operating personnel.

13. The order book of the RR (reactor facility) chief engineer (head).

14. Log of failures and repair take-downs for safety important systems and equipment.

15. In-process documentation (in-process shift log sheet, order book, core fuel loading pattern logs).

16. General and individual quality assurance programs for the RR.

17. Report on the RR physical start-up results.

18. Procedures for experiments in the course of the RR physical start-up.

19. Report on the RR power start-up results.

20. RR certificate

21. Lifetime management program for safety important systems and components.

22. Guidelines on nuclear safety assurance in the course of storage, handling and transportation of fresh and spent nuclear fuel.

 

 

 

 


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