Approved by Resolution of the

Russian Gosatomnadzor dated December, 28, 2000 No. 15

Effectivesince July, 1, 2001

FEDERAL RULES AND REGULATIONS IN THE AREA OF ATOMIC ENERGY USE

REQUIREMENTS TO THE SAFETY ANALYSIS REPORT FOR

SHIP NUCLEAR POWER UNITS

NP-023-2000

These Federal rules and regulations establish the requirements to the safety analysis report for ship nuclear power units which is included into the package of documents substantiating nuclear and radiation safety of these nuclear units and submitted by the operating organization to Gosatomnadzor of Russia in order to obtain the license for design, construction and operation of ship nuclear power units.

The approaches to NPP safety assurance proven by long-term practice and reflected in NPP TSAR with due regard for peculiarities of ship NPUs are used in development of the regulatory document.

The regulatory document is issued for the first time.

The regulatory document is developed by: Russian Research Centre "Kurchatov Institute", Gosatomnadzor of Russia, Scientific and Technical Center for Nuclear and Radiation Safety.

Proposals and comments of the following organizations are taken into account in the regulatory document: Afrikantov Mechanical Engineering Experimental Design Bureau, Russian Research Centre "Kurchatov Institute", Inspectorate for Nuclear-Powered Ships of the Russian Maritime Registry of Shipping, Marine Navigation Department of the Russian Ministry of Transport, Murmansk Shipping Corporation JSC, Iceberg Central Engineering Bureau JSC, Aurora SPA SUE, Krylov State Scientific Center, Central Maritime Fleet Research and Development center subsequent to their discussion at the meetings and development of coordinated solutions.

ABBREVIATIONS

EP - NPU Emergency Protection

PSA - Probabilistic Safety Assessment

GRW - Gaseous Radioactive Wastes

LRW - Liquid Radioactive Wastes

PSS - Protective Safety Systems

CG - Compensation Group

- - Instrumentation

IHCS - Integrated Hardware Control System

IHAT - Integrated Harbour Acceptance Tests

LSS - Localizing Safety Systems

RD - Regulatory Document

NO – Normal Operation

SAR - Safety Analysis Report for Ship NPUs

SSS - Supporting Safety Systems

SFA - Spent Fuel Assembly

SNF - Spent (Irradiated) Nuclear Fuel

ECS - Emergency Cooldown Station

SG - Steam Generator

CW - Commissioning Works

QAP - Quality Assurance Program

PSD - Passive Spray Device

STP - Steam Turbine Plant

RW - Radioactive Wastes

RSb - Radioactive Substances

RP - Reactor Plant

ECCS – Emergency Core Cooling System

SS - Safety System

SRS - Safety-Related System

BAR - Burnable Absorber Rod

I&C - Instrumentation and Control System

PHRS - Passive Heat Removal System

CPS - Control and Protection System

FA - Fuel Assembly

FE - Fuel Element

SRW - Solid Radioactive Wastes

TS - Technical Specifications

CSS – Control Safety System

SNFS - Spent Nuclear Fuel Storage

PCP - Primary Circulation Pump

CCR - Central Control Room

- - Operating Organization

EPS - Electric Power System

NF - Nuclear Fuel

NPU - Nuclear Power Unit

1. GENERAL PROVISIONS

1.1. Purpose of the report

1.1.1. The requirements to the safety analysis report for ship nuclear power units (hereinafter - the SAR) shall be applicable only to NPUs based on double-circuit PWR reactors and establish the structure and content of the NPU safety analysis report submitted to Gosatomnadzor of Russia in order to obtain the license within the package of documents substantiating nuclear and radiation safety.

1.1.2. The SAR shall provide sufficient information for adequate understanding of the nuclear and radiation safety concept used as the basis for the NPU design as well as for adequate understanding of QAPs and the main NPU operation principles.

1.1.3. The SAR shall contain systematic safety analysis for the ship NPU in the course of construction, commissioning and operation in order to confirm absence of any radiation hazard for the ship crew, the public and the environment.

1.1.4. The information presented in the SAR shall comply with the actual NPU state. Presentation of the information shall be compact and shall not contain any ambiguities. Information on compliance with the SAR and other regulatory documents, rules, acts and laws shall not have declarative character. The measurement units corresponding to the design documentation shall be applied in the SAR.

1.2. SAR development procedure

1.2.1. The SAR shall be developed by the ship designer for each particular type of the ship with due regard for the design and NPU peculiarities.

1.2.2. The works for development, compilation and the required adjustment (maintenance) of the SAR shall be performed at all stages of the ship lifecycle.

1.2.3. The SAR developed for the design stage shall be submitted within the package of documents substantiating the NPU safety in the course of application for the ship construction, and the SAR with the adjustments introduced subsequent to the construction and commissioning results shall be submitted in the course of application for the operation license.

The information provided in the SAR with included commissioning results (manufacturing, installation, commissioning works, harbour acceptance tests, physical start-up, IHAT and performance tests) shall comply with the ship as built and ready for operation and shall reflect the actual NPU state.

1.2.4. Any deviations from the NPU design, particularly in the course of the NPU construction, rearrangement, refurbishment shall be assessed from the viewpoint of their impact on the NPU safety and documented by introduction of changes to the relevant SAR sections.

1.3. Requirements for the SAR content, format and maintenance

1.3.1. The content and format of the SAR as well as the procedure for its maintenance (bringing of the SAR information in compliance with the actual NPU state) shall correspond to the SAR requirements. Fulfillement of this condition ensures acceptability of the information contained in the SAR for Gosatomnadzor of Russia and the minimum time limits for its review.

1.3.2. Requirements for the contents

1.3.2.1. As far as possible the SAR content shall be such that no additional review of design or operation materials would be required for safety assessment. References to the design documentation that shall be presented upon additional request of the state safety regulating authority may be given.

1.3.2.2. In case the safety analysis information is based on more detailed works or documents the reference thereof shall be given with indication of the document type, authors or organization, the year of work performance or document issuance, the archive or identification number of the owner.

1.3.2.3. Data redundancy in the SAR shall be avoided. It is recommended to give references to the relevant sections in order to avoid excessive repetitions.

1.3.2.4. Information on the performed calculations and calculation analyses confirming adequacy and completeness of the calculation scope in accordance with the adopted standards, consideration of all factors affecting the results shall be provided. The information shall contain the data sufficient to perform expert calculations in case of necessity (schemes, accepted assumptions, input data, results, their interpretation and conclusions).

All software tools specified in the SAR and assessment of their acceptability shall be briefly described within the scope sufficient for their understanding; their names and validation data shall be provided.

1.3.3. Requirements for the SAR presentation and maintenance

1.3.3.1. Presentation of all sections of the report shall be uniform for all stages of the NPU lifecycle.

The SAR shall be packaged in folders by individual sections or by sections and subsections in case of necessity.

1.3.3.2. The list of abbreviations used in the section shall be provided in the beginning of each section.

The NPU type, the ship design number, the ship name (in case of necessity) and full title of the report and the relevant section (subsection) shall be indicated on the folder.

1.3.3.3. The report shall be issued through the use of computer printing and plotting devices on one or two sides of white paper sheets with A4 format according to GOST 9327.

Graphical information provided in the SAR shall be presented on legible scale and arranged in the text on sheets with format 11 x n or in individual folders on sheets with any format.

1.3.3.4. Numbering of the pages shall be arranged in accordance with the sections representing independent parts. In this case the page number shall consist of the section number and the page number itself and shall be displayed in the top field of the page according to the format "nn-n" or "nn.n-n" for the section.

Digital numbering of paragraphs and sub-paragraphs within sections and subsections shall be used.

1.3.3.5. Any changes shall be introduced into the text of the report by replacement of pages. Introduction of changes by corrections in the text is prohibited.

When individual pages are replaced the reference number of the revision and the date (month, year) of replacement shall be specified on each page in the top right corner on the margins. Replaced pages shall be stored in the end of the relevant sections.

In case of any necessity to increase the number of pages alphabetical notations shall be added, and the relevant notes shall be introduced if this number is reduced.

The record on the reference number of the revision, places and date of the replacement shall be indicated on the first page of the section (subsection) text. The revision history sheet shall be placed in the end of each section.

1.3.3.6. General requirements, approaches, criteria, etc. referred to different parts of the unit shall be presented in the relevant section, and references shall be given in any other sections. The SAR terminology shall comply with the terms used in the regulatory documents.

1.4. Standard structure of the SAR sections

Arrangement of the sections for the relevant safety and safety-related systems and components shall comply with the following standard structure.

1.4.1. Design basis

The purpose of the system, its classification in accordance with the requirements of the federal rules and regulations in the area of atomic energy use shall be specified in this subsection. The list of regulatory documents on safety containing the requirements that the described system shall comply with shall be provided, the principles and criteria used as the basis of the system design shall be specified.

1.4.2. System design

Description of the design and (or) process flow diagram of the system in general, its sub-systems and components (if they perform independent functions) shall be presented. Detailed drawings, figures and schemes illustrating the design of the system and its components, its spatial layout and interfaces with other NPU and ship systems shall be provided. The basic technical characteristics of the system components used for their manufacturing shall be specified. In case any materials inconsistent with the design are used the replacement of materials shall be substantiated with indication of the regulatory documents permitting such replacement.

Selection of the materials shall be substantiated with due regard for the normal operation conditions and any operational occurrences including accidents.

1.4.3. Control and monitoring of the system operation.

Information on automatic, remote and local control and monitoring of the system operation shall be provided. The lists and permissible values for the controlled parameters of the system in all operation modes and in case of shutdown for repair shall be provided; location of the measuring points shall be specified, the control methods shall be described, and information on metrological validation of the applied methods and instrumentation errors shall be presented. The system interfaces with control systems, redundancy of sensors and communication channels shall be described.

1.4.4. Tests and inspections

The scope of the testing program, its objectives, the list of RD and design documentation used as the basis for tests and inspections, the lists of controlled parameter values and the requirements for instrumentation used in the course of testing shall be provided.

1.4.5. Design analysis

The calculation standards, methodologies, input data, descriptions of the computational programs used for calculations, assumptions and restrictions for the calculation schemes, the calculation results and conclusions shall be provided. Information on verification and validation of the calculation programs shall be presented. The scope of information shall be sufficient to perform independent alternative calculations in case of necessity. In case any experiments were performed in order to substantiate safety of the system design the experiment conditions shall be described, their compliance with the design conditions shall be analyzed, the experimental facilities and metrological support shall be described, and interpretation of the testing results and their comparison with the calculation analysis shall be provided.

Description of the system functioning under normal operation conditions, in case of any operational occurrences and accidents, interactions with other systems with due regard for their possible failures and measures for protection of the system against the effects of these failures shall be provided.

Design failures of the system components including any possible human errors shall be listed and analyzed, and impacts of failure consequences (particularly common cause failures) on operability of the system under consideration, any interfaced systems and the NPU and ship safety in general shall be assessed. Failures requiring special consideration in Section 15 of the SAR shall be specified.

Each subsection shall be completed with the conclusions on compliance with the requirements, principles and criteria of the relevant regulatory documents on safety.

Particular content of each subsection may change depending on the system peculiarities.

Individual subsections may be omitted or supplemented with other ones if it is determined by the system peculiarities.

In case any part of the above-mentioned information is presented in other sections the relevant reference shall be given.

As a rule the information shall be presented in the form of tables.

2. GENERAL CHARACTERISTICS OF THE NPU

This section of the SAR shall include the following subsections.

2.1. Basis for the ship design development

The following information on the ship shall be provided in this subsection:

the design development basis, brief information on the published resolutions of the governmental authorities or any other organizations used as the basis for the planned ship construction.

2.2. Ship operation region

Physical, geographical, climatic and meteorological characteristics of the design region for the ship operation and basing including the information on any restrictions of the ship operation with regard to the climatic characteristics, winds, seasons, ice conditions, peculiar natural phenomena (tsunami,

typhoons, cyclones) (if any) shall be provided.

Information on assessment of the soil conditions and engineering properties, slopes, harbours, bays (in order to assess their stability under various natural loads) as well as the plans for potential development of the basing regions (construction of piers, dams, etc.) shall be provided for the basing areas.

Effect of the ship NPU presence in the port shall be analyzed with due regard for the local meteorological conditions, usage of the coastal territories and offshore areas and the population density.

2.3. Brief description of the ship design and its technical characteristics

Brief general information on the ship (within the scope of specifications), namely the lits of basic data on the organizations carrying out design, construction and operation of the ship, manufacturing and installation of the main equipment and safety-related systems shall be provided.

General information on the ship shall contain the following data:

2.3.1. Brief description of the ship class, its main purpose and architectural and structural type

2.3.2. Basic characteristics, the principal ship dimensions and technical characteristics: the maximum length; the length between perpendiculars; the maximum width; the waterline beam; the hull height; draft; displacement; independence with regard to the basic reserves; the crew size; permissible heel and trim difference; resistance to flooding; speed; the propulsion unit type; general layout of the NPU equipment on the ship; fire protection of the NPU (the ship); the containment and shielding devices; radiation safety and radiological control.

2.3.3. Ship hull

General information on the hull, its design and corrosion protection

2.3.4. Hull gear

Information on anchoring, mooring, towing, rescue and cargo handling devices, as well as motors and propellers, steering gear, fire protection and fire detection systems, ventilation and air conditioning systems, drainage and ballast systems, the auxiliary propulsion unit (if any)

2.3.5. Technical description of the design solutions for the NPU

2.3.5.1. Materials on the NPU design organization, its contractors and manufactures of the main NPU equipment (the reactor, the nuclear core, the SG, the emergency propulsion boiler, heat exchangers, electric motors, the turbine set, generators, turbine generators, diesel generators, pumps, compressors, valves, boiler, refrigeration and evaporation plants, condensers, auxiliary boiler plants, heaters, control systems, instrumentation)

2.3.5.2. Information on the configuration of the hardware control and automation systems

2.3.5.3. Information on configuration and location of the electric power systems, control rooms, lighting, communication and alarm systems and any other EPSs installed on the ship; possibility for receipt of energy media (direct and alternate current power supply from the shore (the ship), bidistillate, diesel fuel, compressed air) and supply to the shore (the ship); normal and emergency lighting, welding equipment, navigation, communication, alarm and control means

2.3.5.4. Brief description of the NPU

2.3.5.5. Information on the NPU configuration and location

2.3.5.6. Configuration and brief characteristics of the RP

2.3.5.7. Information on the normal NPU operation modes including description of the NPU commissioning; the NPU operation for the intended purpose (at the constant power level, in transient modes, in case of any power level changes); the NPU bringing into the hot standby mode with subsequent switching to the normal operation mode or to the cooldown mode; the NPU operation in case of any failures and malfunctions, particularly ones resulting in power limitations as well as in case of any risks for the ship safety; power ascension after unintended fast power shedding or spurious activation of emergency protection

2.3.6. Brief description of the radiation protection including:

- the main criteria of radiation protection for the crew, the public and the environment including the offshore area, brief description of the means aimed to ensure the minimum exposure levels;

- the personnel exposure dose limits;

- the means aimed to prevent LRW discharge into the sea;

- the means for RW collection, storage and disposal;

- radiation and contamination levels for each zone of the ship and any access restrictions for these zones due to such levels;

- RW handling rules and procedures;

- rules and procedures for access into the controlled and supervised areas;

- description of the biological protection (information on the protective structures, sources to be protected, location and purpose of the protective structures as well as their dimensions and materials, assessment of the structures according to the calculation or testing data);

- comparative data on the radiation level expected or permitted in the course of operation and in case of design basis accidents.

2.3.7. Brief description of the radiological control system containing the information on the design principles, location, type, sensitivity and measurement ranges, applied sensors, information display and alarm methods, reliability and durability of the radiological control system

2.3.8. The list of initiating events, design basis and beyond design basis accidents analyzed in the SAR

2.3.8.1. Ship accidents (accidents for the conditions at sea and at a port):

- collision with water ingress into the power and auxiliary compartments;

- stranding, girting, drowning in shallow water, drowning in deep water, fire in the CCR and the power compartment, including the engine room, the electric power compartment, the reactor compartment, the IHCS equipment rooms

2.3.8.2. NPU accidents

The following shall be considered:

a) loss of the primary circuit integrity;

b) unauthorized reactivity changes including unauthorized movement of the most efficient

reactivity control devices with the speed permitted for the control system and the drive design; cold water injection into the SG and the reactor, malfunction of the feed valve; unauthorized activation of the emergency cooldown system;

c) heat removal malfunctions, including:

- malfunctions in the primary circuit due to partial or complete loss of forced circulation in the primary circuit, erroneous activation of the pump, pressure drop in the primary circuit;

- loss of the primary circuit integrity - a small leak in the primary circuit, leakage in the SG piping system, inter-circuit leaks in the primary circuit equipment, the reactor cover leakage, pressure increase in the primary circuit;

- malfunctions in the secondary circuit due to any rupture of the condensate feeding pipeline, the main steal pipeline rupture, pressure increase in the steam pipeline, closure of the main shut-off valve upstream of the turbine, loss of steam removal from the RP, isolation of the main condenser, loss of cooling water supply to the main condenser, shutdown of the condensate and feeding pumps, feedwater flow rate decrease up to complete stop;

d) malfunctions in the power supply system due to loss of power supply on one hull side, malfunction of the turbine generator on one hull side, loss of the control and protection system power supply, loss of power supply for the actuator of any CPS control device, short-term or long-term complete blackout of the NPU;

e) violation of habitability conditions in the CCR and the NPU rooms due to inoperability of the conditioning and ventilation system.

2.3.8.3. Possible course of events after a failure or an accident, recommendations for accident management with regard to the main areas for mitigation of consequences

Description of the results for each analyzed event shall contain the input data for the analysis, basic assumptions used to perform the calculations, errors of the calculations, the containment leakage value and efficiency of the ventilation system (adsorption and filtration), the assigned automatic actions or the required actions of the personnel, the time period after events within which the urgent measures and arrangements for confinement of the event consequences shall be implemented.

2.3.8.4. Reliability of the equipment and other components

Information on reliability of the equipment and the SRS components including the list (range) of reliability parameters for each type of equipment subject to reliability substantiation, results of calculation (calculation and experimental) substantiation of the reliability parameters, conclusions on compliance of the reliability parameters with the RD requirements, results of qualitative reliability analysis, assessment of uncertainties for the reliability analysis results, assessment of any potential impact of incompletely considered factors on the calculations, the list of components with significant contribution to reliability of the systems, references to the applied calculation methodologies and programs, characteristics of the input data on reliability shall be provided. The information shall be presented in the form of tables for each type of equipment.

2.3.8.5. Probabilistic safety assessment

Information on the PSA results including characteristics of the initial reliability database, the list of considered initiating events and its substantiation and the following information shall be provided: performed qualitative and quantitative analysis, reliability of the systems (results of interaction between the systems shall be presented in the form of tables), the applied models of failure trees and event trees, including the information on the success criteria used for the main systems, consideration of common cause failures, actions and errors of the personnel, external events, assessment of sensitivity

and uncertainty, the final PSA results.

The information on balance of the design and any changes introduced into it on the basis of the PSA in order to achieve this balance shall be provided, the main contributors to any severe accident risk and distribution of their relative contributions shall be specified.

2.3.9. Construction conditions

2.3.9.1. The subsection shall contain brief description of the shipyard location area, climatic conditions including the temperature of ambient air, the outer water, the ice conditions, depths in the moorage areas, current velocity, navigation conditions, geological, hydrological and seismotectonic characteristics, seismic activity in the region; information on the population density in the construction area, characteristics of any extreme natural impacts, location of airfields and rocket airfields shall be provided.

2.3.9.2. The subsection shall contain description of the production process, requirements for the technological level of the production infrastructure and the management structure, including management of work preparation at the shipbuilding plant, availability of production and testing equipment, the necessary technical and process control system, the basic qualification requirements for the personnel.

2.3.9.3. General criteria for safety assurance in case of any fire on the ship in all operation modes as well as in case of design basis and beyond design basis accidents shall be provided in the subsection; in this case the fire shall be considered as an initiating event or consequence of an initiating event with due regard for the single failure principle. The fire safety assurance concept shall be formulated, and its criteria shall be specified.

Multiple barriers, the optimal correlation of passive and active protection, adequate redundancy of the safety system channels, their physical separation, etc. shall be substantiated; the fire consequences shall be assessed with due regard for any possible failures of fire extinguishing units.

Systematic approach to fire safety assurance and stage-by-stage planning of the arrangements aimed to improve fire protection training of the ship crew shall be demonstrated; fire hazard shall be analyzed for different sections of the ship with indication of the hazardous factors and quantitative assessment of fire load. The ship zoning principle (division into fire zones and compartments) shall be described.

Impossibility of simultaneous loss of control from the CCR and the ECS due to a fire as well as impossibility to lose more than one channel ensuring performance of the safety function shall be substantiated.

In case the ventilation systems are used to ensure confinement of the fire within one fire compartment, and components of these systems are considered as SSS or LSS elements the consequences of any single failure of these components for general safety assurance shall be analyzed, and redundancy of these components shall be substantiated in case of necessity.

In case active fire extinguishing systems belong to supporting safety systems their arrangement principle, reliability level, capability of these systems to withstand single failures and analysis of any extreme impacts on the fire detection and extinguishing means shall be specified.

Consequences of any spurious actuation of fire extinguishing units shall be analyzed with regard to impact on the safety-related equipment from the viewpoint of general safety assurance.

2.3.9.4. The subsection shall include the construction stages, particularly the lists of potentially hazardous works at each stage, arrangement of transition from one stage to another, expected radiation situation in case of any accident, probability assessment for the RSb releases into the atmosphere,

potential exposure doses for the shipyard workers and the public, as well as the list of measures for nuclear and radiation safety assurance, analysis of the work performance compliance with the requirements of effective regulatory documents on nuclear and radiation safety, the arrangements for the public protection, the list of process documentation required to support the NPU construction process.

2.3.10. NPU operating conditions

2.3.10.1. The following information shall be presented in the subsection:

- the limit conditions for the ship operation in case of design basis accidents (anticipated operational occurrences when further operation of the NPU with certain restrictions defined in the design is permitted);

- examination and inspection of the NPU technical condition, the time limits for scheduled inspections (frequency and scope, including testing);

- procedures for approval and modification of operational guidelines, orders, decisions on the service life extension;

- the procedure for the crew staffing, the personnel number and qualification;

- procedures and guidelines defining arrangement of manegement under normal operation conditions and in case of any operational occurrences and accidents.

2.3.11. Information on physical protection of the NPU

2.3.11.1. Administrative, engineering and technical measures of physical protection with analysis of compliance with the requirements of the effective standards and rules with regard to physical protection assurance, analysis of the ship NPU and efficiency of its physical protection system, the list of documents on arrangement and functioning of the ship NPU physical protection system shall be provided.

2.3.11.2. Analysis of the ship NPU and efficiency of its physical protection system shall be provided.

2.3.11.3. The list of documents on arrangement and functioning of the ship NPU physical protection system shall be provided.

2.3.12. Information on the standards and rules applied at the ship lifecycle stages

The information may be presented in a separate document with the relevant security label in accordance with the legislation of the Russian Federation.

The section shall include the list of documentation used to substantiate the NPU safety.

3. GENERAL PROVISIONS AND APPROACHES TO THE NPU SAFETY ASSURANCE

3.1. Ship NPU safety assurance concept

The list of regulatory documents considered in the NPU safety analysis shall be provided in this subsection; it should be demonstrated that the NPU safety assurance concept is implemented in the design. For this purpose the following shall be provided:

- information on application of the inherent RP safety principle in the design and engineering solutions for its implementation;

- substantiation of the defense-in-depth principle implementation through the use of the physical barrier system and the multi-level system of technical and administrative measures aimed to protect the barriers and to maintain their efficiency;

- substantiation of adequacy of the means aimed to ensure safe residual heat removal from the nuclear core.

3.2. Nuclear and radiation safety assurance

The nuclear and radiation safety objectives shall be specified in this subsection, and the systems used to achieve these objectives shall be demonstrated.

3.2.1. Nuclear safety assurance

Nuclear safety shall be substantiated with regard to the following areas:

- Maintenance of control over chain nuclear reaction in the reactor nuclear core.

The extent for usage of the reactor inherent safety properties in order to ensure nuclear safety shall be demonstrated. Information on the reactivity balance for all possible operational states, emergency situations and design basis accidents shall be provided, the possibility of any positive reactivity effect occurrence in case of accidents shall be analyzed, and their potential consequences shall be assessed. Efficiency, reliability and fast response of the reactor EP shall be substantiated.

- Heat removal from the reactor nuclear core.

The principal scheme and substantiation of the reactor nuclear core cooling under normal operation conditions and in case of any operational occurrences including design basis and beyond design basis accidents shall be provided, the passivity degree shall be assessed for the heat removal systems adopted in the design.

- Prevention of local criticality during refueling, transportation and storage of NF.

Adequacy of the measures for prevention of local criticality in the course of NF handling at all stages of the NPU lifecycle shall be substantiated.

3.2.2. Radiation safety assurance

Engineering features and administrative measures aimed to ensure protection of the personnel, the public and the environment against unacceptable radiation exposure shall be described. It should be demonstrated that application of the proposed protection means and implementation of the protective measures are justified by practice and do not result in exceedance of the established dose limit, prevent any unreasonable irradiation, and the existing radiation exposure is maintained at the lowest reasonably achievable level with due regard for the economic and social factors. Efficiency of the protection systems and their sufficiency to guarantee insignificant damage to the personnel health, the public and the environment shall be demonstrated.

3.3. The NPU safety systems and the basic principles of their arrangement

The subsection shall include the information on safety systems, particularly on the solutions provided in the NPU in order to ensure the required protection level, the main functions performed by the safety systems, the SS arrangement schemes and their compliance with the requirements of standards and rules, implementation of the basic arrangement principles in the safety systems (single failure, priority, safe failure, conservative approach, irreversibility and controllability of the functions, practical evaluation, diversity, multiple channels, physical separation), resistance of the safety systems to common cause failures, arrangements aimed to ensure performance of the SS functions under

external impacts and in case of human errors, considered beyond design basis accidents, arrangements for their management, experience in design, construction, testing, commissioning, operation and decommissioning confirming adequacy of the administrative and technical measures for the NPU safety assurance.

3.4. Classification of the ship NPU systems and components

It should be demonstrated that the systems and components are divided into safety classes (SC1, SC2, SC3 and SC4) depending on their importance for the NPU safety in order to ensure safety of the ship NPU and to minimize radiation exposure for the crew and the public as well as radioactive contamination of the environment in all operational and emergency states of the ship NPU, and the design requirements for the materials, manufacturing, testing and operation are established in accordance with these safety classes.

3.5. Initial states and zoning of the ship rooms

It should be demonstrated that the initial states to be considered in the NPU design are classified in accordance with the probability of their occurrence within the range of constant manifestation to extremely rare; the ship is divided into the restricted areas (the controlled access area, the supervised area and the uncontrolled access area) depending on the potential hazard of radiation contamination.

The list of rooms occupied with the NPU systems and components shall be provided.

3.6. Design dose limits and exposure levels

It should be demonstrated that the main dose limits, the permissible exposure levels for the personnel and the public and the standards for RSb release, discharge and content in the environment adopted in the design comply with the requirements of the regulatory documents.

3.7. Design conditions, principles and criteria for the NPU

It shall be demonstrated in the SAR that the NPU can be operated under the maritime conditions, the NPU design provides for periodic inspections and tests of the safety systems in the course of operation without any safety level reduction, the impact loads on the NPU components caused by all external impacts, etc. considered in the design are taken into account, the safety systems retain their operability in case of the ship heel of up to 30°, or roll of up to 45°, or trim by bow or stern of up to 10°, the radiation exposure for the public and the crew does not exceed the safe levels defined in the Radiation Safety Standards (NRB-99).

4. REACTOR PLANT

Information on the main technical characteristics of the RP as the source of thermal energy, the data on the number and type of the reactors, the main thermotechnical equipment, supporting systems, the SG type, the coolant volume thermal change compensation system shall be provided.

4.1. Reactor

Information and analysis sufficient to substantiate safe operation of the nuclear reactor within its design service life under normal operation conditions and in case of any operational occurrences including design basis accidents, as well as the information required to analyze the accidents and to obtain the results presented in Section 15 of the SAR shall be provided.

Information and analysis provided in this section shall be based on the basic design materials for the reactor, the nuclear core, the nuclear core components (fuel elements, fuel assemblies, the control devices of the control and protection system, the BARs, etc.), the reactor internals and other systems

important for the RP safety assurance.

Design of the reactor any any systems and components located inside the reactor shall be described in this section, their classification in accordance with the regulatory documents defining the design criteria and principles shall be presented. Process flow diagrams of the described systems and components with the references to the relevant design documentation shall be provided. Interaction of the described structures, their mutual impact, assembly and dismantling conditions shall be understood from the description.

As fas as possible description of the safety-related systems and components located inside the reactor shall be presented in accordance with the standard structure given in Subsection 1.4 of the SAR.

4.2. Nuclear core

4.2.1. Purpose and design basis

Purpose and design basis for each nuclear core loaded into the ship reactors within its service life and its components as well as the basic characteristics of the nuclear core shall be described in this subsection. The list of regulatory documents containing the requirements the nuclear core and its components shall comply with shall be presented.

Compliance of the nuclear core and its components with the requirements of the effective RD on safety as well as the requirements of the terms of reference for the nuclear core shall be analyzed.

4.2.1.1. Materials

The list of regulatory documents governing the requirements for the applied materials shall be provided.

4.2.1.1.1. Structural materials

The following information shall be provided:

- validation of the structural materials for the nuclear core components or analysis of their operation experience under the same conditions;

- any changes of the mechanical, thermal and physical properties depending on the neutron fluence, temperature, time, initial state of the metal;

- permissible stresses and deformations with due regard for cyclic loading;

- corrosion resistance with due regard for the medium chemical composition, initial state of the metal, temperature, time, neutron fluence, steam content, peculiarities of heat removal. Evidence of resistance to inter-crystalline corrosion, stress corrosion, pitting corrosion and stability under the impact of decontamination solutions shall be presented;

- radiation creep, swelling, shape changes depending on temperature, neutron fluence, time of irradiation, initial state of the metal;

- compatibility of materials (chemical and metallurgic interaction) in case of contact with each other under normal operation conditions, any operational occurrences and accidents.

4.2.1.1.2. Welding

The following information shall be provided:

- applied types of welding in accordance with the list of regulatory documents governing the requirements for welding;

- operation experience for weld joints or their testing under similar conditions;

- differences of mechanical and corrosion properties of weld joints in comparison with base metal under normal operation conditions, in case of any operational occurrences and accidents.

4.2.1.1.3. Nuclear fuel

The following information shall be provided:

- chemical composition, dimensions, enrichment, density, loading (particularly with regard to fissionable isotopes), NR distribution irregularities, control methods, metrological validation of the control means;

- permissible burn-up depth (analysis of operation experience);

- volumetric, linear and phase changes, changes of density, heat capacity, thermal conductivity, mechanical properties and gas emission in the nuclear fuel depending on burn-up, temperature, thermal cycles, neutron fluence;

- compatibility with the cladding material, mass transfer, impact of the fission products on mechanical strength of the structural materials;

- behavior in emergency situations (loss of fuel element leak-tightness, contact with the coolant, temperature increase);

- experimental substantiation of the fuel data (the list of calculations and reports).

4.2.1.1.4. Absorbing materials

The following information shall be provided:

- chemical composition, dimensions (particle size), enrichment with regard to absorbing isotopes, density, control methods, metrological validation of the control methods;

- permissible burn-up of absorbing isotopes (analysis of operation experience);

- volumetric, linear and phase changes of the poison, gas emission, changes in mechanical properties, heat emission, thermal conductivity, density depending on time, burn-up, temperature, neutron fluence;

- compatibility with the cladding (coating) material;

- behavior in emergency situations (loss of fuel element leak-tightness, contact with the coolant, temperature increase);

- the list of calculations and reports on experimental substantiation of the information on absorbing materials.

4.2.2. Design description and drawings

Design description and general drawings or assembly drawings of the nuclear core components (fuel assemblies, fuel elements, control devices of the control and protection systems, BARs, neutron sources, etc.) demonstrating mutual arrangement, dimensions, lattice pitch for fuel elements and fuel assemblies, allowances, clearances, sealing techniques for the components, their fastening and

orientation shall be provided.

The nuclear core fuel loading pattern and information on the NF loading shall be provided.

4.2.3. Compliance with the requirements of General Safety Provisions for Ship NPUs and Safety Rules for Ship NPUs for the nuclear core shall be analyzed.

The design limits (operational, safe operation limits, limits established for design basis accidents) shall be provided. The permissible measurement errors for control of the parameters or calculation errors shall be specified for all established limits.

The operational characteristics of the nuclear core under any maneuver modes shall be provided.

Thermotechnical, neutron and physical and strength analyses of the nuclear core safety calculations shall take into account:

- vibration characteristics of the nuclear core components in the coolant stream, presence of any resonance;

- internal and external pressure on the fuel element claddings, the BARs, stress-strain state of the claddings;

- corrosive and erosive damages, generation of hydrogen and hydrides, corrosion depositions on the claddings;

- causes and consequences of any non-conformities between the reactor power and the coolant parameters;

- changes in dimensions and shape of the nuclear core components due to irradiation, creep, temperature variations, blocked CPS devices;

- interaction between the fuel and the cladding (mechanical, chemical) particularly in case of the fuel element leak-tightness loss and contact of the fuel with the coolant;

- stability of protective coatings;

- probability and expected scale of leak-tightness loss in any fuel elements and BARs under normal operation conditions, in case of emergency situations and design basis accidents;

- energy emission and potential chemical reactions in case of leak-tightness loss in any fuel elements;

- energy emission and the resulting pressure impulse in case of the fuel element cladding breakage, the fuel element filling with water and ingress of fuel debris into the coolant;

- any potential failures of fuel elements, BARs in case of any coolant circulation reduction or loss.

4.2.4. Control, monitoring and testing

The list of controlled parameters of the nuclear core and its components shall be provided and substantiated with indication of the integrity criteria for the claddings of fuel elements and fuel assemblies, frequency of control, measurement ranges for the parameters, permissible measurement errors, dynamic characteristics of the measuring channels. Brief description of the CPS, the diagnostic system as well as characteristics of the parameters (setpoints) for the EP activation shall be provided. More detailed information shall be presented in Section 7 of the SAR, and the relevant reference shall be given.

Testing programs and methods for the nuclear core and its components, the non-destructive and destructive control and testing methods confirming the design characteristics of the nuclear core components shall be described; the list of regulatory documents defining the requirements for the control and testing scope and methods shall be presented. The programs of incoming control for the nuclear core components, the inter-departmental testing of the nuclear core and the acceptance certificate of the inter-departmental commission, the list of nuclear-hazardous works with the nuclear core and its components shall be provided.

4.2.5. Quality assurance

The main provisions of the nuclear core quality assurance program and the program implementation report shall be presented at the design stage, or reference shall be given to Section 17 of the SAR which shall contain detailed information on the quality program.

4.2.6. Nuclear core design assessment

Safety assessment in accordance with the criteria adopted for the nuclear core shall be provided. In case of failure to comply with any criteria or insufficiency of the provided evidence of compliance technical and (or) administrative measures to be implemented in order to comply with these criteria shall be specified, and it shall be confirmed that these criteria will be met after implementation of these measures; the schedule for implementation of these measures or submittal of the missing information shall be presented.

4.3. Neutron and physical substantiation of the nuclear core design

4.3.1. Design basis

Design neutron and physical characteristics of the nuclear core loaded into the reactor (-s) and the reactivity control systems shall be described, and restrictions with regard to neutron and physical characteristics and reactivity control shall be specified, such as:

- design duration of the NF campaign and NF burn-up;

- the maximum reactivity margin;

- negative reactivity feedback and reactivity coefficients;

- movement velocities of the reactivity compensation devices;

- the limit controlled reactivity insertion rate;

- sub-criticality margins after emergency shutdown of the reactor in the course of the campaign.

4.3.2. Description of the neutron and physical characteristics

4.3.2.1. Peculiarities of the neutron and physical characteristics shall be listed, described or illustrated for characteristic moments of the campaign.

They shall include the following aspects:

- NF profiling in the fuel assemblies and the nuclear core;

- distribution of the burnable absorber.

4.3.2.2. Power density distribution

Quantitative information on the design power density distributions for normal operations shall be provided, including distributions inside standard fuel assemblies, axial distributions, general radial distributions among the fuel assemblies in the nuclear core as well as power density distribution within the nuclear core volume.

In order to make sure that the standard distributions expected under NO conditions are described to the full extent, and impact of all possible parameters on these distributions is taken into account both standard (normal) and limit (maximum) power density distributions related to standard and limit manifestations of the parameters, power, flow rate, flow rate distribution irregularity, position of the rods, the campaign moment (fuel burn-up and potential burn-up distributions), the burnable absorber and accumulation of poisons shall be reflected in sufficient details.

Characteristics values of errors or uncertainties that can be related to design power density distributions shall be specified.

The design power density variation coefficients (with regard to the form and numerical values) occurring under the limit manifestations of the steady states as well as for the initial conditions and the transient modes shall be provided.

4.3.2.3. Reactivity coefficients

Complete quantitative information on the design reactivity coefficients shall be provided.

The information shall be presented mainly in the graphical format and shall cover the entire ranges of parameters (density, temperature, pressure, void content, power) from the cold start-up to the limit values used in the safety analyses.

Sufficient information shall be provided to illustrate normal and limit values of the parameters related to the operational and emergency states, the campaign moment, the position of the burnable absorber rods, power density distribution, density of the moderator, etc. Potential uncertainties of the calculation results shall be specified, and the experimental results confirming the calculations and the adopted uncertainty values shall be presented. Tests and inspections provided for the reactor shall be described. In case the coefficient limits are of particular importance (for example, for positive reactivity coefficients with regard to the moderator temperature) the reactor testing for verification of these limits shall be described in the maximum details.

4.3.2.4. Requirements for reactivity control

Tables and results related to the nuclear core reactivity balances shall be provided for the beginning and the end of the campaign and also for the intermediate moments of the campaign in case of necessity. Description of the factors affecting the reactivity and depending on different operational states shall be provided, including:

- purpose of the CGs, the EP groups, their expected and minimal permissible efficiency;

- efficiency of burnable absorber;

- acceptability of any "blockage" for the movable reactivity compensation devices;

- temperature disturbances of the moderator and fuel as well as any possible void disturbances;

- burn-up (slags);

- xenon and samarium poisoning;

- permissible depths of the reactivity compensation device insertion into the nuclear core and their acceptable mismatch.

The minimum required and expected sub-criticality margin for the promptly shut-down reactor shall be presented and substantiated for different moments of the campaign with due regard for uncertainties of such margin and experimental verification in operating reactors.

Methods and restrictions for control under normal operation conditions shall be described in detail reflecting such aspects as:

- movement of CGs and EP groups affecting power density distribution;

- potential changes in the coolant flow rate or temperature.

The following shall be described:

- start-up from the cold, hot and the most poisoned state;

- the load following mode and compensation of non-steady xenon poisoning;

- impact on volumetric power density distributions and burn-up distributions.

4.3.2.5. CG location and efficiency

Complete information on the CG location shall be provided. Detailed information on their segregation into groups, sequence and degrees of their withdrawal from the nuclear core, any substantiated restrictions for their positions depending on the power level, the campaign moment or any other parameters shall be included.

4.3.2.6. Reactor sub-criticality in the course of refueling

The maximum K value shall be established for the reactor with

eff.refueling, and the conditions for non-exceedance of this value shall be substantiated.

4.3.2.7. Stability

Presence or absence of stability in the reactor operation shall be demonstrated. In case of stability statement the confirming data for similar reactors shall be provided.

Criteria used to define whether the reactor operation will be stable shall be specified.

In case any instability or limited stability is expected the measures aimed to detect and eliminate any exceedance of safety limits shall be provided.

4.3.2.8. Neutron irradiation of the pressure vessel

Distributions of the neutron flux and fluence in the nuclear core, at its boundaries and on the walls of the reactor pressure vessel within the reactor service life specified in the design shall be provided.

4.3.3. Analytical methods

Analytical methods used in the neutron and physical calculations including the methods for determination of criticality, reactivity coefficients and burn-up effects shall be described. The applied computer codes (programs) shall be described in detail with indication of the code name and type, its validation based on critical experiments or confirming predictions for the operating reactors. Description of the codes shall include description of the methods for determination of such parameters as neutron cross-sections. Accuracy of the analytical methods shall be assessed.

4.3.4. Changes

Any changes of the design for each nuclear core loaded into the NPU reactors within its service life, calculation methodologies, data or information related to the parameters important for neutron and physical calculations shall be listed together with the affecting parameters. Peculiarities and impact of the changes shall be considered in the relevant SAR sections.

4.4. Thermohydraulic calculations

4.4.1. Input data for the calculations

The input data for the thermohydraulic calculations of the reactor with regard for each type of nuclear cores loaded into the NPU reactor (-s) within its service life shall be provided including geometry of the coolant flow channels, criterial correlations for the heat transfer and hydraulic processes, characteristics of the coolant, particularly two-phase flow parameters, maximum temperature of the fuel and fuel element claddings, the burnout ratio, characteristics of the clearance between NF and the fuel element cladding (if any) as a function of burn-up, circulation velocity, the coolant distribution, the coolant volume, flow resistance and flow stability, restrictions for unstable modes, criteria of integrity for the claddings of fuel elements and fuel assemblies.

4.4.2. Description of thermohydraulic calculations for the nuclear core

The thermohydraulic calculation program for the reactor shall be described including the following information:

4.4.2.1. Comparison of the main thermohydraulic calculation parameters for the reactor with the data obtained on similar operating reactors This comparison shall include, for example, the input coolant temperature, temperature of the fuel, maximum and average linear heat rates, the burnout ratios, the coolant velocities, surface thermal flows, the specific output, geometry of the heating surface and the circulation zone.

4.4.2.2. Critical heat flux coefficients for the nuclear core point with the maximum neutron flux (or with the least favorable heat removal conditions) at the rated power and design overloads The applied calculation methodologies, comparison with the results obtained through the use of other methodologies or by experiments shall be provided.

4.4.2.3. Linear heat rate

Average and maximum values of linear heat rates shall be specified for all design nuclear core areas.

4.4.2.4. Steam content

Curves demonstrating expected radial and axial distributions of steam generation in the nuclear core shall be presented. The average steam content value for the nuclear core and the maximum steam content value for any section of the nuclear core shall be specified.

4.4.2.5. Coolant flow distribution

Distribution of the coolant flow in the nuclear core and the applied throttling as well as the input data for throttling calculations with due regard for any power distribution changes in the course of the nuclear core campaign shall be described and substantiated.

4.4.2.6. Flow resistance

Pressure differential values in the nuclear core and flow resistance under normal operation conditions and in case of any accidents shall be specified with due regard for the data presented in par. 4.4.2.5.

4.4.2.7. Heat transfer coefficients

Calculation methodologies and physical characteristics used to determine the heat transfer coefficients shall be specified and substantiated.

4.4.2.8. Temperature effects in transient operation modes

The nuclear core capability to withstand heat impact in case of any anticipated transient operation modes shall be assessed.

4.4.2.9. Calculation uncertainties

Uncertainties related to assessment of the limit and boundary conditions for thermal and hydraulic calculations shall be demonstrated.

4.4.3. Description of the thermohydraulic calculations for the reactor coolant circulation system

Thermohydraulic calculations shall be described for the nuclear reactor coolant circulation system. The description shall include the following information:

4.4.3.1. Brief table of thermal and hydraulic characteristics

The table summarizing thermal and hydraulic characteristics of the reactor coolant circulation system shall be provided.

4.4.4. Analysis of calculations

Analysis of thermal and hydraulic calculations for the nuclear reactor and the coolant circulation system shall be presented including the following information:

4.4.4.1. Critical heat flux

The following shall be specified:

- the ratio between the critical heat flux and the critical power;

- experimental data substantiating the above-mentioned ratio. Special attention shall be paid to substantiation of the methodology intended to take into account the impact of the coolant flow distribution, its mixing and spatial power distribution.

4.4.4.2. Nuclear core hydraulics

Analysis of the nuclear core hydraulic scheme shall include the following:

- description of the model trial results and assessment of their applicability for the calculated nuclear core with due regard for various circulation paths through the reactor and distribution of the coolant at the nuclear core inlet;

- analysis of applicability of the empirical correlations used in the calculations within the entire range of the planned reactor operation modes;

- consideration of the effect of partial or complete separation of the coolant circulation loops.

4.4.4.3. Natural coolant circulation mode

Conditions for formation and maintenance of the natural coolant circulation mode shall be specified.

4.4.4.4. Analytical methods

Analytical methods and data used to determine the flow velocity in the reactor coolant circulation system shall be described in detail. The description shall include classic patterns and empirical dependencies covering both single-phase and double-phase circulation modes (if any are provided). Uncertainties of the calculations and the resulting uncertainty in determination of the flow velocity in the reactor coolant system shall be assessed.

Analytical techniques used to calculate thermal and hydraulic characteristics of the nuclear core including assessment of uncertainties shall be described. This description shall take into account hydraulic instability, the neutron flux intensity, presence of the most heat-stressed channels, impact of contaminations and depositions as well as operation with one or several isolated circulation loops.

Information on validation of the calculation programs shall be presented.

4.5. Tests and inspections

Testing and inspection programs and methods used to confirm design characteristics of the nuclear core and the nuclear reactor coolant circulation system within the entire nuclear core campaign shall be presented. References to the relevant subsections of Section 12 of the SAR may be given.

The following shall be provided:

- requirements for the instrumentation and hardware to be used for monitoring and measurement of the thermal and hydraulic parameters important for safety assurance;

- requirements for the instrumentation and hardware with the sensors located inside the nuclear core and in the circuit lines intended to confirm the expected distributions of specific output and temperature of fuel elements, the moderator and the energy characteristics of the reactor;

- characteristics of the hardware used in the RP for monitoring of vibrations and any loosening in the fastening of the equipment parts;

- methodologies for determination of impermissible vibrations and any loosening in the fastening of the equipment parts.

4.6. Materials used for the reactor manufacturing

4.6.1. Materials of the reactor pressure vessel and cover

The data confirming compliance of the reactor vessel materials, manufacturing and control methods with the requirements of standards and rules shall be presented.

4.6.1.1. Technical specifications for the materials

Materials used to manufacture the reactor vessel and cover as well as materials of the equipment contacting with the reactor vessel shall be listed. Technical specifications for the materials shall be indicated.

Criteria for selection of the materials shall be specified, and compliance with these criteria shall be substantiated.

4.6.1.2. Manufacturing technology

The principal technology for manufacturing of the vessel and cover components shall be described with indication of the heat treatment regimes and welding type as well as the relevant regulatory documents.

In case any non-standardized or special techniques are used it should be demonstrated in detail that their application will not affect integrity of the reactor vessel and cover.

4.6.1.3. Non-destructive control methods

The methods for detection of any surface and internal defects shall be described in detail, references to the methodologies and the quality control program shall be given with indication of the relevant regulatory documents.

4.6.1.4. Special control methods for carbon and austenitic stainless steels

Requirements of par. 4.6.1.4 are similar to the requirements of par. 4.6.1.3. In case it is recommended to select the RD, the control method and scope out of several alternative variants the selected variant shall be substantiated.

4.6.1.5. Brittle failure resistance

Requirements of par. 4.6.1.5 are similar to the requirements of par. 4.6.1.4. Besides, the temperature variation curve shall be presented for the reactor vessel and cover.

4.6.1.6. Material state monitoring in the course of operation

Requirements of par. 4.6.1.6 are similar to the requirements of par. 4.6.1.5. Besides, the following shall be provided:

- description of the control program for surveillance specimen, characteristics of the specimen, their set and the proposed sampling schedule;

- the scheme of the specimen location in the container and location of the containers in the reactor, the container fastening technique, representativeness of the specimen location shall be substantiated (from the viewpoint of the neutron flux fluence and temperature);

- expected impact of irradiation on the material characteristics based on the validation tests of the material (for example, shift of the critical brittle point), strength calculations, graphical characteristics.

4.6.1.7. Fasteners of the reactor vessel and cover

Materials and design of the fasteners for the reactor vessel and cover shall be described.

Non-destructive control operations in the course of manufacturing shall be specified with the reference to the quality control program, type, scope and frequency of control in the course of operation.

4.6.2. Design limits for pressure and temperature

The pressure and temperature limits adopted in the design for normal operation modes, operational occurrences, design basis accidents and hydraulic testing shall be substantiated.

4.6.2.1. Limit values

Pressure and temperature limits shall be specified for the following conditions:

- preliminary hydraulic factory testing of the primary circuit components;

- operational testing of the primary circuit for leak-tightness and strength;

- normal operation including warm-up and cooldown rates.

The temperature and pressure values established in the design shall be specified in the SAR at the design stage.

Results of strength testing for the materials and the temperature and pressure limits based on the obtained characteristics shall be included into the SAR, and the expected irradiation impact shall be also demonstrated. The input data used for prediction shall be described.

4.6.3. Integrity of the reactor pressure vessel and cover

Information on integrity of the pressure vessel and cover not presented in any other sections shall be specified. In this case probability of the reactor vessel and cover breakage (with the reference to the performed analysis) and factors supporting maintenance of its integrity as well as the reactor designer and manufacturer and their experience level shall be specified.

It should be demonstrated that the reactor pressure vessel and cover can withstand static and dynamic loads under normal operation conditions, in case of any operational occurrences and design basis accidents without any breakage within the entire service life.

4.6.3.1. Design

Design principles and criteria adopted in the course of the design development shall be provided. Safety class in accordance with the Basic Provisions on Safety Assurance for Ship NPUs shall be specified.

Brief description of the design, its sketch with indication of the components and materials shall be provided; peculiarities of the design and manufacturing methods shall be specified separately. Regulatory documents used in the design development shall be listed, compliance with the design principles and criteria shall be substantiated.

4.6.3.2. Manufacturing methods

The adopted manufacturing methods shall be specified, compliance with the requirements of standards and rules shall be demonstrated. Experience in operation of pressure vessels and covers manufactured through the use of these methods shall be described.

4.6.3.3. Requirements for control

Design requirements for control of the pressure vessel and cover integrity shall be specified; in case the design requirements are stated by the design organization they shall be substantiated. The control methods adopted by the designer in addition to the ones prescribed in the regulatory documents shall be described. The methods for registration of the initial inspection results for the pressure vessel and cover shall be described.

4.6.3.4. Transportation and installation

The means for protection of the reactor vessel and cover in the course of transportation used to protect it against any environmental impacts, particularly against corrosion and damage, peculiarities of transportation, the permissible transportation means shall be specified.

The handling techniques, the installation scheme with indication of the main operations including erection of the pressure vessel on the supports shall be provided.

4.6.3.5. Design limits

The design limits ensuring safety of the pressure vessel and cover shall be specified for the normal operation conditions and any operational occurrences. Assurance of the pressure vessel and cover integrity under the most stressed modes shall be substantiated.

The main stages of sealing and unsealing of the main reactor joint with the cover and other pressurized detachable joints shall be specified with indication of the measures to ensure strength and tightness of the joints (the assembly procedure, the tightening force, the control methods, etc.).

4.6.3.6. Control in the course of operation

The procedure and scope of inspections for the reactor vessel and cover shall be described.

Information on the applied control means, their characteristics and experience of application at similar facilities confirming their acceptability shall be provided.

Measures aimed to ensure adequacy and comparability of control in various periods of operation (including the incoming and post-installation control) shall be described.

4.6.4. Structural materials of the CPS drives

The CPS drives shall include all external electromechanical devices and units up to the points of their connections to the CPS control devices. The list of materials and their technical specifications for each part of the CPS drives, information on the mechanical properties with due regard for their operation modes, substantiation of the possibility to use these materials, information on their validation or operation experience under similar conditions shall be provided.

4.6.5. Materials of the reactor internals

4.6.5.1. The list of materials and their technical specifications for the main parts of the reactor internals, information on their mechanical properties with due regard for their operation conditions, substantiation of the possibility to use these materials, information on their validation shall be provided.

4.6.5.2. Weld joints

Information on the requirements for welding of the reactor internals and the list of regulatory documents determining these requirements shall be provided.

4.6.5.3. Non-destructive control

Non-destructive control methods and means for the materials, parts and structures of the reactor internals shall be described, and the list of applied regulatory documents shall be provided.

4.6.5.4. Manufacturing and processing of the reactor internal components

Requirements for mechanical and thermal processing of the materials, the relevant control program and the list of applied regulatory documents shall be provided.

4.7. Design functioning of the reactivity control systems

4.7.1. Information on the CPS

The information shall include general drawings of the CPS drives, kinematic schemes of the drive operation, the layout drawings for the CPS drives, schemes of cooling pipelines, instrumentation and

hardware, descriptions and characteristics of the units, auxiliary equipment and their hydraulic systems. As far as possible the description of the CPS drives shall comply with the standard structure given in Subsection 1.4.

It should be demonstrated that all safety-related equipment is sufficiently protected against common cause failures.

4.7.2. Testing and control of the CPS drives

The functional testing programs including checks of insertion and withdrawal for all CG and EP groups, thermal and hydraulic tests simulating operation and emergency modes the CPS is to withstand shall be provided.

The CW programs shall be provided; objectives and methods of testing as well as the system acceptance criteria shall be substantiated.

4.7.3. Information on location of the CPS drives

The section shall include the layout schemes for the drives and equipment in plan and profile view.

4.7.4. Operation analysis for the reactivity control systems

Functional characteristics of the joint reactivity control systems in case of any accidents shall be assessed.

5. PRIMARY CIRCUIT SYSTEM AND THE ASSOCIATED SYSTEMS

Issues with regard to safe functioning of the primary circuit and maintenance of its integrity under normal operation conditions, in case of any operational occurrences, emergency situations and after postulated initiating events not related to the primary circuit leak-tightness loss shall be reviewed in this section. The leak-tight primary circuit represents the next barrier limiting propagation of radioactive substances after the fuel element claddings.

The primary circuit provides heat transfer from the nuclear core to the working medium of the secondary circuit and includes the reactor, the PCP, the SG, pipelines (if any) connecting the above-mentioned components.

Credibility of the information presented in the section shall be confirmed by the safety analysis results provided in the SAR. The information shall be complete enough, and all necessary safety analyses shall be performed.

Information on the primary circuit components and the associated systems demonstrating compliance of the primary circuit system with the RD requirements, including the Basic Provisions on Safety Assurance for Ship NPUs and Nuclear Safety Rules for Ship NPUs shall be provided.

The list of the basic design documents used as the basis for development of this section shall be presented.

Components and systems included into the primary circuit and the associated systems shall be considered in this section:

- the main coolant circulation line;

- systems (or parts of systems) related to the main circulation line within the primary circuit pressure boundary:

- the reactor vessel and cover;

- systems ensuring normal functioning of the main circulation line of the primary circuit:

- pressurizing;

- normal reactor cooldown;

- emergency cooldown and emergency cooling of the nuclear core;

- high pressure gas;

- the secondary, third and fourth circuits;

- overpressure prevention for the primary circuit and the SG;

- coolant purification;

- auxiliary systems:

- the primary circuit make-up;

- air removal, sampling and drainage;

- loading and unloading of the active filter sorbents;

- liquid poison supply;

- detection of any leaks in the primary circuit equipment;

-the primary circuit valves;

- fasteners and expansion joints of the long-distance pipelines.

Spacing elements (supports, shock absorbers, limit stops, etc.) between the primary circuit components and the ship (foundation) structures shall be considered within each system.

Note. The complete set of systems and components shall be defined in accordance with the particular design.

5.1. Brief description

5.1.1. The primary circuit and associated systems

Performance of the main safety function of the primary circuit - heat removal from the nuclear core by sufficient amount of the coolant with proper quality under normal operation conditions, in case of any operational occurrences and design basis accidents in compliance with the operation limits and safety limits, particularly with regard to fuel element cladding damage limits, shall be demonstrated in this section, and the list of postulated initiating events shall be provided.

The following shall be specified:

- information on the design and safety analysis results for the primary circuit systems and components;

- description and purpose of the primary circuit, its main components and the associated systems. Components performing independent functions as well as safety functions of each system and

component shall be defined in the description;

- tables of design and working (operating) characteristics;

- safety criteria and principles adopted in the design;

- references to the design sheets of the primary circuit systems and components;

- information on the calculations performed in the course of design, the list of experimental works and analysis of the experimental results;

- references to other sections of the SAR containing more detailed requirements for individual systems and components of the primary circuit;

- description of all elements installed on the pipelines and equipment in order to accommodate any loads occurring due to sea disturbance, heeling, trim differences and other external impacts;

- information on the fact that the design provides for submittal of information to the operator;

- any operational occurrences in the primary circuit;

- reaching of the operation limits and (or) safe operation limits or current values of the parameters.

The following shall be demonstrated in this subsection:

- the design provides for monitoring of the temperature and pressure in the primary circuit and the level in the pressurizers under normal operation conditions, in case of any operational occurrences, emergency situations and design basis accidents;

- all systems and components of the primary circuit are designed with due regard for the possibility to withstand unfavorable ambient conditions (pressure, temperature, humidity, radiation, rolling, impact loads, etc.) occurring under normal operation conditions, in case of any operational occurrences, emergency situations, design basis accidents and their consequences within the entire service life;

- the possibility for the radioactive coolant drainage in absence (presence) of any dead zones, water filling and gas removal from the system is provided. It should be demonstrated that the primary circuit is designed in such a way so that to ensure access to the equipment for decontamination, inspections, maintenance and repair works, and the exposure doses for the personnel do not exceed the limits established in the design.

5.1.2. Process flow diagram

The process flow diagram of the primary circuit with indication of the primary circuit boundaries and all main components, the operating pressure, temperatures, flow rates and coolant volume in the steady plant operation mode at full-rated power shall be provided. All systems connected to the primary circuit and the ways of their connection and isolation from the primary circuit shall be indicated on the flow diagram. It is important for the systems with non-radioactive media and the systems with the working pressure lower than the one in the primary circuit.

The pipeline routing within the reactor compartment shall be presented in isometric view.

5.1.3. High pressure gas system

Values of the basic parameters used to activate the main and redundant groups, their characteristics and efficiency under normal operation conditions, in case of any operational occurrences

and emergency situations shall be provided. Redundancy of the system components and gas reserves as well as the state of the system during design basis accidents in case of gas loss shall be specified. The information shall be presented in accordance with the scheme given in par. 5.2.2.

5.1.4. Instrumentation and hardware scheme

The instrumentation and hardware diagram of the primary circuit and the associated non-isolable systems within the pressure boundary of the primary circuit shall be provided. The list of instrumentation for measurement of pressure, temperature, flow rate, level, chemical composition of gas and water as well as for monitoring of the water flow rate and leak-tightness of the primary circuit with indication of the accuracy class for the instruments shall be provided.

5.1.5. General drawings of the primary circuit system

The general drawings indicating mutual arrangement of the equipment and basic dimensions of the primary circuit components in relation to the supporting and surrounding structures and demonstrating the possibility for maintenance and inspection shall be provided. In case the design provides for any biological protection it shall be indicated.

5.2. Integrity (strength and tightness) of the primary circuit pressure boundaries

Any measures provided in the design in order to ensure strength and leak-tightness of the primary circuit equipment and pipelines shall be substantiated in this subsection.

It should be demonstrated that all equipment and pipelines withstand static and dynamic loads without any breakage.

5.2.1. Compliance with the standards and rules

The table demonstrating compliance with the requirements of standards and rules adherence to which is supervised by Gosatomnadzor of Russia shall be provided.

5.2.2. Overpressure protection of the primary circuit

The list of components performing the overpressure protection functions in the primary circuit shall be provided.

All measures and techniques for protection of the primary circuit systems against overpressure above the design limits under normal operation conditions, in case of any operational occurrences, emergency situations and design basis accidents shall be listed in this subsection.

References shall be given to any other SAR sections where individual systems and components protecting the primary circuit against destruction shall be described. Information on the individual systems shall be presented in accordance with the scheme given in the section "General provisions".

5.2.2.1. Technical arrangements adopted to reduce the probability of any pipeline rupture and equipment breakage with separation of parts should be demonstrated in the safety analysis.

In this case the following shall be presented:

5.2.2.1.1. Pipeline breakage criteria

Data on any potential pipeline rupture stress concentration points as well as the zones with potential risk of any damage to the adjacent safety-related equipment.

For low-temperature conditions the design data shall be provided to confirm that the pressure in the primary circuit components at low temperatures (below the operating temperature) is limited to the values eliminating brittle failure, or the pressure corresponds to the level of stresses permissible for this temperature range.

5.2.2.1.2. Analysis of pipeline rupture consequences

Analysis results for consequences of any pipeline ruptures shall be presented in this subsection with due regard for the impacts on the adjacent equipment:

- temperature, pressure, water jets and steam on the broken pipeline and any units connected to it;

- loads on the adjacent equipment and pipelines from kinetic energy of jets caused by water and steam releases;

- reactive loads causing vibration and whipping in the damaged pipes;

- damages caused by missiles;

- humidity and radiation;

- flooding of safety-related equipment.

In case the "leak before break" concept is applied the pipelines it is applied to shall be specified, and reference to the document substantiating its application shall be given.

5.2.2.1.3. Protection against the consequences of pipeline breakages

Methods used in the design for physical separation of the pipelines and restrictions of movements shall be described in order to prove that:

- rupture of any primary circuit pipeline does not result in rupture of any other pipeline required for mitigation of the accident consequences;

- rupture of any pipeline not belonging to the primary circuit does not result in any accidents with loss of coolant;

- rupture of any primary circuit pipeline does not result in the containment destruction;

- coolant discharge does not prevent any works in the control rooms and does not impair any systems used to eliminate the accident consequences.

5.2.2.2. The following shall be specified:

- description of the overpressure protection techniques for the primary circuit;

- results of calculations for these loads and the resulting stresses;

- results of the analysis demonstrating the effect of any changes in the operation modes, parameters and operating characteristics of the equipment on the system characteristics.

5.2.2.3. Tests and inspections

Tests and inspections to be performed prior to commencement of operation, during the RP start-up in order to confirm the functional characteristics and in the course of operation shall be specified.

5.2.3. Primary circuit materials

Data confirming compliance of the materials, manufacturing and control methods for the primary circuit pressure zone components with the requirements of standards and rules applicable in nuclear shipbuilding shall be provided.

5.2.3.1. Technical specifications for the materials

The list of technical specifications for ferritic and austenitic stainless steels, non-ferrous metals particularly titanium ones (if any are applied) used for manufacturing of the primary circuit components including fasteners as well as welding and surfacing materials shall be provided.

The ways to consider the material properties listed below and significantly affecting the pressure boundary integrity assurance in selection of the primary circuit materials shall be demonstrated:

- chemical compatibility with the coolant;

- compatibility with the material of any components contacting with the pressure circuit (heat insulation, supports, coatings, parts of sealing units, etc.);

- cyclic strength, creep-rupture strength and yield;

- corrosion (with due regard for stress corrosion), corrosive and cyclic and erosion characteristics;

- radiation damage (for steels subjected to neutron exposure);

- resistance to cracking;

- brittle failure resistance;

- fabricability;

- activation under irradiation;

- behavior in emergency situations.

Information on presence of any chemical elements with adverse effect on the operational characteristics in the applied steels (for example, cobalt content in nickel-containing steels; copper, nickel and phosphorus content in the reactor vessel steel; carbon, sulfur, phosphorus and silicon content in carbon steels, etc.) as well as any measures aimed to restrict application of such materials shall be provided.

5.2.3.2. Compatibility of structural materials with the primary circuit coolant

The following information related to compatibility of the primary circuit coolant with structural materials and external insulation of the pressure zone shall be provided:

- - chemical composition of the primary circuit coolant with the reference to the relevant regulatory document and indication of any expected chemical composition changes in various reactor operation modes, the permissible content of chlorides, fluoride compounds, oxygen, hydrogen and soluble corrosion products;

- compatibility of the structural materials with the primary circuit coolant including the list of structural materials contacting with the primary circuit coolant and description of compatibility of the materials with the coolant, impurities and radiolysis products they can contact with. In case any non-metal materials contact with the primary circuit coolant compatibility of these materials with the coolant shall be described;

- compatibility of the structural materials with external heat insulation of the primary circuit

including the list of the primary circuit structural materials with heat insulation and description of their compatibility with external heat insulation especially in case of any coolant leakage. Information on non-metal heat insulation of austenitic stainless steel shall be provided in order to demonstrate whether concentration of chlorides, fluoride compounds, sodium and silicates in the heat insulation will be within the acceptable limits; these limits shall be substantiated.

5.2.3.3. Manufacturing and processing of components made of carbon steels

Information on manufacturing and processing of carbon and low-alloy steels shall be provided, particularly:

- process for manufacturing of semi-finished products and items in accordance with the supporting documentation;

- description of non-destructive control operations for all components within the primary circuit pressure zone. Reference to the QAP shall be given.

5.2.3.4. Manufacturing and processing of austenitic stainless steels

The following information on manufacturing and processing of austenitic stainless steels used in the primary circuit components shall be specified:

- peculiarities of the procedure for processing of the items (forging, welding, heat treatment) preventing crack formation due to stress corrosion as well as limitations with regard to ferritic phase. The control methods used in the course of manufacturing and enabling to detect stress corrosion in the items shall be specified;

- control of the processes in order to reduce the contact with any media capable to cause stress corrosion. Measures for protection of the component surfaces against contaminations and damages promoting corrosion cracking (from the manufacturing stage till the installation completion);

- characteristics and mechanical properties of cold-deformed austenitic stainless steels for the primary circuit components and the permissible deformation degree;

- measures aimed to prevent hot cracking in the course of welding and assembly. The requirements for welding materials shall be specified. Compliance of the welding technique including repair of weld seams and control (particularly qualification of welding operators) with the requirements of standards and rules applicable in nuclear shipbuilding shall be demonstrated;

- non-destructive control operations for the primary circuit components; reference to the quality control program shall be given.

5.2.3.5. Interface with the secondary circuit

The following shall be specified:

- technically possible amount of coolant flowing to the secondary circuit in case of any circuit-to-circuit leakage in the SG;

- minimum volume of water and maximum volume of steam in the SG under normal operation conditions.

5.2.3.6. In-service inspections and testing of the primary circuit

The program of in-service inspections and testing for the primary circuit components shall be described. The description shall contain the following:

- boundaries of the systems subject to control including supports and fastening elements;

- location of the systems and components taking into account access for control purposes;

- control methods and techniques;

- control frequency;

- requirements of the in-service inspection program;

- methods to assess the control results;

- frequency and procedure for hydraulic testing (for strength and leak-tightness).

Peculiarities of in-service inspections and testing for individual primary circuit components shall be specified, and references to the relevant design sections shall be given.

5.2.4. Detection of leakages through the primary circuit pressure boundaries

The leakage detection system shall be described in accordance with the scheme specified in par. 5.2.2.

The applied leakage detection methods, sensitivity, response time and also reliability of the instruments and equipment shall be described; the minimum leakage value that can be detected through the use of the applied methods shall be specified.

Besides, the systems (techniques) used for alarm and serving as indirect indicators of leakage shall be presented.

Combinations of the techniques (systems) provided in the design for detection of the leakage point shall be demonstrated.

The program for processing of sensor signals enabling the operator to obtain reliable information on location and intensity of any leakages shall be described.

Testing methods for the primary circuit leakage detection systems shall be described.

5.3. Components of the primary circuit and the associated systems

Information on the components included into the primary circuit pressure boundaries and the associated systems shall be provided in this subsection. It should be sufficient to assess their impact on safety of the entire ship and shall include purpose of the components and systems, design criteria, characteristics and description of the design, assessment of compliance with the established design criteria.

The component (system) counterpart with well-known operation experience shall be specified; any variations from the counterpart and the reasons to introduce them shall be indicated.

In case any component (or system) is completely taken from any other plants, or commercial products are used it shall be demonstrated that their technical characteristics, operation modes and conditions comply with the requirements for the RP under consideration.

In case any component (or system) is newly developed its necessity shall be substantiated.

The QAPs applicable to the particular component (or system) shall be described. Impact of any damages and failures of the components on the RP safety shall be demonstrated, and any failures with

the consequences requiring special analysis shall be indicated.

As the number of the primary circuit components and the associated systems can vary for different RP types the applicant shall define the set of these components and systems for the particular NPU type and subsections for each component and system depending on their peculiarities. Anyway, the calculation substantiation, description, the required testing, decontamination and inspections shall be specified for each component or system associated with the primary circuit, apart from the above-mentioned information, and assessment of the components or the system in general shall be provided. Peculiarities of maintenance associated with the radiation level shall be taken into account.

Requirements for the particular information to be presented in the SAR in addition to the information specified in this subsection are given below. This information shall reflect the peculiarities of individual primary circuit components.

5.3.1. Primary circuit circulation pumps

The scope of presented information shall include the data on the PCP design and characteristics as well as the description of the PCP auxiliary systems, their characteristics, design criteria and substantiation of their fulfillment. Brief description of the PCP instrumentation and hardware, auxiliary systems with the list of protections and interlocks restricting the PCP operating conditions shall be provided.

5.3.2. Steam generators

The information shall be presented in the form given in par. 5.2.2. Besides, the SG characteristics shall include design limits of radioactivity level in the SG secondary circuit under normal operation conditions and substantiation of these limits.

Radiological consequences of any breakage of heat exchanging tubes, the SG header and other design basis accidents related to the primary-to-secondary circuit leakage shall be considered.

The design criteria aimed to prevent unacceptable damage of the SG heat exchanging tubes (due to vibration, corrosion damage, etc.) shall be demonstrated, and their fulfillment in the design shall be substantiated.

The following shall be provided in the calculation substantiation:

- design conditions and assumptions, the list of analyzed operation modes (from among normal operation modes, operational occurrences and emergency modes) determinant for strength assessment of the heat exchanging tubes and their embedding points in the headers;

- results of calculations and experiments confirming that the adopted stress intensity level ensures reliable operation of the SG;

- evidence of integrity maintenance for the heat exchanging tubes and the SG headers in case of design basis accidents with large-scale leaks (ruptures) of the primary and secondary circuit pipelines outside the SG;

- heat exchange surface reserve.

5.3.2.1. SG materials

Information on selection of the materials shall be provided with due regard for the SG peculiarities and its manufacturing technology affecting the requirements for the materials (for example, presence of the steam and water separation zone, temperature variations, design and embedding technique for the heat exchanging tubes, etc.). The way to consider these peculiarities in selection of the materials (for

example, the necessity to improve characteristics of the materials with regard to crack resistance, corrosion resistance) shall be demonstrated.

Information on peculiarities of the SG design (if any) capable to affect changes of the material properties in the course of operation shall be provided.

Compatibility of the SG materials with the primary and secondary circuit coolant shall be substantiated. The manufacturing technology for the main SG components shall be described in brief. Techniques for the heat exchange surface cleaning in the course of manufacturing and cleanness control methods shall be described. Material selection for the heat exchanging tubes shall be substantiated, and requirements for the surface condition, heat treatment, corrosion resistance and other parameters important to ensure operability of the tubes shall be specified.

Description of the SG transportation methods, measures provided in the design in order to prevent damage of any SG components in the course of transportation and installation, the heat exchange surface preservation necessity and technique, preservation and cleanness control for the inner surface in the course of storage, installation and final assembly at the shipyard shall be presented. The SG installation procedure shall be described in brief.

5.3.2.2. SG control and maintenance in the course of operation

The arrangements provided in the SG design in order to monitor state of its components in the course of operation shall be described.

Detailed description of the control techniques and methods for the heat exchanging tubes shall be presented. Labor inputs for the control and the related dose commitments shall be assessed.

Description of the equipment used for control, the control accuracy, the registration methods, the assessment criteria, frequency of control, measures to be taken in case of any defect detection, particularly the techniques for elimination of any defects in heat exchanging tubes shall be provided.

The most important SG maintenance procedures in the course of operation shall be described, including the cleaning technique for heat exchanging tubes aimed to recover their heat transfer capability and decontamination; characteristics of the secondary circuit water chemistry regime and the measures for its maintenance provided in the design shall be specified. WCR restrictions preventing from further operation of the SG in case of their violation shall be specified.

5.3.3. Pipelines containing the primary circuit coolant

Information on the pipelines operated under the primary circuit pressure (the non-isolable part of the primary circuit) shall be provided.

The information shall be presented in the form given in par. 5.2.2.

The relevant references to the information on the criteria, methods and materials specified in Section 5.2.3 shall be given in the description of the pipelines.

Arrangements for monitoring of the factors promoting stainless steel cracking due to stress corrosion shall be demonstrated in the general description.

In case the "leak before break" concept is used in the design the regulatory documents substantiating its application shall be provided.

5.3.4. Isolating valves for the secondary and third circuit equipment

Description and technical characteristics of the isolating valves of the secondary and third circuit aimed to isolate the equipment from the primary circuit in case of any circuit-to-circuit leakage shall be

provided.

5.3.5. Emergency cooldown system

All methods (systems) used in the design for residual heat removal from the nuclear core shall be listed with indication of their functions.

Similar information shall be provided for the systems removing heat from the primary circuit in case of accidents.

The information shall be presented in accordance with the scheme for systems and equipment given in par. 5.2.2.

5.3.6. Pressurizer

The information shall be presented in accordance with the scheme of systems and equipment given in par. 5.2.2.

5.3.7. Valves

Information on shut-off, isolating and control valves of the systems associated with the primary circuit shall be provided. The information shall be presented in accordance with the scheme given in par. 5.2.2.

5.3.8. Support structures of the main components

Sketches and brief description of the support structures for the reactor, the SG and the PCP with indication of the loads they are designed for shall be provided.

6. STEAM TURBINE PLANT

The information on the STP shall be provided in this section, including:

- configuration of the STP, its main equipment and systems. Part of the STP equipment and systems shall be considered as SRS. The STP systems and components ensure emergency cooldown of the NPU as well as emergency cooling of the nuclear core (the reactor flushing) and the primary circuit make-up and shall be considered as PSSs and SSSs.

Information on the STP devices and systems including the condenser, the main and auxiliary steam systems, venting, blow-off, aspiration and steam-air mixture removal systems, the condensate feeding system, the feedwater receiving and pumping system, the outer water circulation system, the oil system and any other systems affecting the RP safety shall be provided.

6.1. Turbine set

6.1.1. Design basis

6.1.1.1. Purpose and functions

The turbine set functions under normal operation conditions, in case of any operational occurrences, emergency situations and accidents, its impact on the RP shall be demonstrated; the turbine set safety class shall be also defined and substantiated, and the list of regulatory documents used to develop it shall be provided.

6.1.1.2. Design modes and input data

Requirements for the turbine set cyclic load capability with indication of the permissible number of start-ups within the service life (cold start-up, hot start-up, scheduled and unscheduled shutdowns, load reduction to idle run, load reduction to the lowest limit of the adjustment range with subsequent loading); design duration of start-ups in various thermal conditions from steam supply to the turbine set up to the rated load; the adjustment range of automatic power variation; deviation of the rotor rotation rate within the adjustment range and under emergency conditions shall be specified.

The rated characteristics of all turbine set operation modes including the start-up and shutdown conditions shall be described.

6.1.1.3. Process system and design option

Functioning of the steam-condensate cycle scheme shall be described with indication of the adopted engineering solutions for arrangement of the main components, as well as purpose of the components and systems; brief description of the operation modes, safety functions performed by the components shall be provided.

6.1.1.4. Layout requirements

Layout of the turbine set shall be substantiated, location and orientation of the turbine set, the areas of potential missile ejection from the turbine set rotor, location of the equipment in the turbine hall shall be presented on drawings.

6.1.2. Design of the turbine set

The basic design principles and requirements for the layout solutions shall be substantiated.

6.1.2.1. Structure of the turbine set

The process flow diagram and the design of the turbine set particularly the overspeed control system for the turbine set rotor shall be substantiated, including substantiation of redundancy for the control and monitoring devices, the applied overspeed controller type. Design of the turbine set, the type of control valves, vibration characteristics of the blades and techniques for their attachment to the rotor rims, support and thrust bearings, vibration characteristics of the rotor in assembly shall be described. The STP layout drawings (plans and profiles), the steam-condensate cycle scheme, schemes of oil supply, control, protection and alarm with indication of the control parameters and their links with the CCR shall be attached to the descriptions.

The lists of initiating events shall be provided.

6.1.2.2. Components

Safety classification of the turbine set components shall be substantiated, strength characteristics of the turbine setting components and the following information shall be provided:

- programs used to assess the blading strength;

- information on brittle strength of the rotor, calculation results for the turbine rotor stage disks;

- information on strength characteristics (maximum tangential and radial stresses and areas of their localization).

Calculation programs shall be described.

6.1.2.3. Materials used

Information on the materials used to manufacture the parts of the turbine set, the rotor, disks,

working blades, mechanical properties, tensile strength characteristics of the rotor material shall be included.

6.1.2.4. Protection against impermissible overpressure

Selected means for the turbine protection against impermissible overpressure shall be substantiated in brief, and their operation principles shall be described.

6.1.2.5. Overspeed protection

The overspeed protection system for the turbine, the redundancy methods, reliability assessment for the components and the entire system and the control and testing procedure shall be substantiated.

6.1.2.6. STP shutdown

Techniques and conditions for the STP shutdown shall be substantiated, the turbine set state after shutdown shall be described.

6.1.3. Control and monitoring of the STP operation

6.1.3.1. Protections and interlocks

Protections and interlocks affecting emergency protection of the reactor, preventive protection (if any) and emergency RP power reduction shall be described.

6.1.3.2. Measurement points

Reference to the process flow diagram with indication of all measurement points shall be given in the description of measurement points.

6.1.3.3. Safe operation limits and conditions

Safe operation conditions of the turbine set causing emergency reactor power shedding shall be substantiated.

6.1.4. Tests and inspections

Arrangements aimed to ensure quality of the turbine set and its equipment in the course of manufacturing, construction and installation shall be described.

The scope and methods for the incoming control, pre-operational commissioning tests and their metrological support shall be provided.

6.1.5. Design analysis

6.1.5.1. Reliability parameters

Reliability parameters for the turbine set and its equipment, results of qualitative and quantitative reliability analysis shall be substantiated. Calculation of the reliability parameters shall be integrated for the STP with due regard for the associated systems.

In case any experiments were performed in order to substantiate reliability brief information on their results shall be provided.

The scope of information on the equipment reliability parameters and the calculation programs shall be sufficient to perform independent alternative calculations.

6.1.5.2. Normal operation

All normal operation modes of the turbine set (start-up, power operation and shutdown) shall be substantiated in brief with indication of the factors affecting the RP operation. In particular, effects of sudden load shedding and potential transient processes shall be reflected while demonstrating the operation of the turbine set control system and overspeed protection system.

The basic actions of the personnel in the course of the turbine set functioning in case of any operational occurrences, emergency situations and accidents shall be described.

6.1.5.3. Turbine set functioning in case of operational occurrences

Information on the qualitative analysis of any potential failures of the turbine set and its systems shall be provided.

Options of the turbine set operability recovery due to redundancy of the systems or its temporary operation with the equipment switched off shall be demonstrated.

Functioning of the STP in case of any operational occurrences and abnormal operation of the systems interfaced with the turbine plant shall be considered.

6.1.5.4. Analysis of operational occurrence consequences

Analysis of failures, initiating events and accidents resulting in increase or decrease (up to complete loss) of the feedwater and steam flow rate, pressure increase and decrease in the secondary circuit, the feedwater temperature increase or decrease as well as the list of initiating events resulting in emergency situations and accidents shall be provided.

Functioning of the STP in case of any emergency situations and accidents shall be demonstrated with due regard for the operation of its components.

6.1.5.5. Functioning of the STP under external impacts

The state (operation or shutdown) of the turbine set under all external impacts (rolling, ice ingress into the circulation lines, stranding) shall be reflected, the level of external impacts when the turbine set shall be shut down should be substantiated.

6.1.5.6. Design assessment

The design assessment based on the qualitative analyses, information on any possible experiments and quantitative reliability parameters of the turbine set shall be provided.

Compliance with the RD requirements for safety and technical specifications for supply shall be demonstrated.

6.2. Requirements for the associated systems of the turbine set

The systems functionally related to the turbine set, particularly the main steam pipeline system with venting to the main condenser, the exhaust steam system, the condensate and feeding systems, the steam pipeline blow-off system, the sealing aspiration and steam-air mixture removal systems, the feedwater receiving and pumping system, the oil system, the oil receiving, pumping and discharge system, the water chemistry maintenance system of the secondary circuit, the outer water cooling system shall be listed. Functional purposes of the systems and their interfaces with other systems shall be specified. Descriptions of the systems performing any SRS functions shall be presented in accordance with the structure provided in Subsection 1.4. In this case impact on the STP operation and failures affecting the RP shall be demonstrated for each system.

Systems that do not perform any SRS functions shall be described in accordance with the following structure:

- design of the system, purpose and description, analysis of the system design; it should be demonstrated that the system is not required for safety assurance.

6.3. Operability substantiation for the system components

Strength, stability and operability of the system components (pipelines, pumps, gate valves, main valves, safety and relief valves) under any natural and human-induced impacts shall be substantiated.

In accordance with classification of each system components and loads the results of calculations shall be provided in order to confirm strength, stability and operability of these components in all design operation modes.

In case any commercial equipment is used deviations of the design operation mode characteristics from the ones given in the technical specifications for supply shall be provided.

7. CONTROL AND MONITORING

The NPU control and monitoring systems and means under normal operation conditions and in case of any operational occurrences including accidents when protection of the process equipment, the ship crew, the public and the environment against any potential radioactive releases is required shall be considered in this section.

The requirements for the information presented in this section shall be applicable to the NPU safety-related systems and components performing control and monitoring functions for the purpose of safety assurance.

Requirements for the information on the control aspects related to safety substantiation for the reactor control systems under normal operation conditions and in case of any operational occurrences including design basis accidents, the reactor emergency protection systems, the systems for submittal of safety-related information to the operator, safety-related control and monitoring systems and other normal operation systems not capable of affecting the NPU safety in case of any failure shall be considered.

Requirements for the information shall also apply to the safety aspects related to arrangement of the NPU control by the operating personnel and their safety-related functions.

The information shall be presented within the scope and with the particularization degree required to substantiate technical and administrative safety assurance solutions adopted in the design.

The requirements shall be applied to both systems and means performing the control and monitoring functions through the use of usual standard control and automation devices and the automated control systems using control computers, information and computational systems and microprocessor equipment.

7.1. Main control and monitoring objectives

7.1.1. Definition of safety-related control and monitoring systems and means

All safety-related control and monitoring systems and means as well as alarm and communication systems performing control and monitoring functions in order to achieve the following objectives shall

be listed:

- safety assurance under normal NPU operation conditions;

- prevention of any deviations from safe operation limits and conditions;

- prevention of any operational occurrences and design basis accidents;

- mitigation of any beyond design basis accident consequences;

- return of the NPU into the controlled state in case of any operational occurrences and design basis accidents;

- arrangement of the personnel management and annunciation under normal operation conditions and in case of accidents.

Names and designations of all monitoring and measuring means shall be specified in accordance with the design documentation and TS.

Classification of these systems and means in accordance with their purpose and nature of the performed functions shall be provided; newly developed systems and means as well as any applied standard commercial and field-proven systems and means shall be specified.

7.1.2. Basic safety principles and criteria

All input data for the analysis, principles, criteria, safety standards and rules, regulations, guidelines and any other documents taken into account for design of the systems used to achieve the objectives specified in par. 7.1.1 shall be listed.

Compliance of the systems with the requirements of safety standards and rules shall be demonstrated.

Compliance with the requirements of any other applied regulatory documents shall be described.

7.2. Systems supporting normal operation of the NPU

7.2.1. Instrumentation and control system of the NPU

7.2.1.1. Purpose and design basis

Information on the purpose of the system and components (instrumentation and hardware, controls, sensors, transducers, etc.), safety principles and criteria used as the basis for the design shall be provided.

Functions of the system (components) shall be defined, and criteria for performance of these functions shall be specified.

7.2.1.2. Description

Information containing description of the NPU I&C system, data on its configuration, basic technical characteristics, description of the system operation principle under normal operation conditions and in case of any operational occurrences and accidents with due regard for interaction with other systems and any equipment related to it shall be provided.

Information on the NPU I&C system and its components shall be provided, including:

- systems and means ensuring remote, automated and (or) automatic control of the systems under normal NPU operation conditions;

- means that ensure monitoring of the parameters characterizing the NPU functioning in all possible ranges of normal operation conditions and presentation of any information thereof as well as information on any deviations from normal operation conditions;

- the information support systems for the operator, including the system for in-process submittal of consolidated information on the current NPU safety state to the personnel;

- hardware stae diagnostic systems in various operation modes;

- diagnostic means for the NPU I&C hardware and software and the radiation situation monitoring systems.

Information on the I&C system and its components shall contain the data on the configuration, the main technical characteristics, location, schemes of the systems and components, description of their operation principle under normal operation conditions and in case of any operational occurrences and design basis accidents with due regard for interaction between the systems and components and the interfaced equipment.

The input information used for safety analysis particularly the methods for assessment and control of the reliability parameters at different stages of the system development and operation shall be provided.

Information on power supply, protection against external impacts, the systems maintaining the ambient parameters for the I&C equipment and the personnel shall be provided.

Special attention shall be paid to substantiation of application of any materials, systems and devices, new equipment, control and monitoring methods as well as substantiation of application of any imported and custom-made devices, their comparison with the counterparts at operating nuclear-powered ships.

Schemes of information flows, coding systems, figures, schemes, diagrams, graphs, tables required to substantiate the adopted engineering and technological solutions for safety assurance shall be provided and described. The NPU I&C components not required for safety assurance shall be specified.

7.2.1.3. Commissioning works

The list of potentially hazardous works in the course of the I&C system maintenance, administrative and technical measures to be implemented for performance of these works shall be provided.

The operation limits and conditions for the I&C commissioning stage shall be substantiated. In case the final requirements for the operation limits and conditions, sequence and scope of the commissioning works are established at the detailed design stage the relevant information shall be provided in the SAR.

Special attention shall be paid to the methods of operability checks for the control and monitoring systems and hardware, their integrated setup, diagnostics and documenting of their characteristics, acceptance criteria and their substantiation.

Information on comparison with the similar administrative and technical solutions for design of the NPU I&C system and its components with due regard for trial and testing of counterparts and prototypes shall be provided.

7.2.1.4. Maintenance

Operation limits and conditions for the NPU I&C system used as the framework to prevent any deviations from the safe NPU operation limits and conditions shall be substantiated.

Special attention shall be paid to substantiation of the solutions related to diagnostics and regular state monitoring for the NPU I&C system and its constituent parts, devices and components, their periodic inspections and functional tests, recording and documenting of malfunctions and failures as well as training of the personnel.

The information presented in this subsection shall contain the input data for analysis of any safety impact of the NPU I&C system maintetance. Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.

7.2.1.5. Safety analysis

The analysis results for the character and impact of the control and monitoring system failures not representing initiating events of any accidents, analysis and nature of accidents demonstrating the degree of compliance with the design criteria and the requirements of the safety rules and regulations shall be provided.

The information shall contain the analysis results for reactions of the systems and devices on any external and internal impacts (fires, flooding, electromagnetic interference, short circuits of the primary power supply network, etc.), reactions of the systems on any potential failures and malfunctions (loss of the insulation quality, voltage drops and induced noise, spurious actuations, loss of control, etc.), as well as results of the quantitative reliability analysis, results of the stability analysis for the control and adjustment circuits and their safety impact, types of reading devices, number of the reading channels, the parameter measuring ranges in these channels, accuracy and location of the instruments, calculation adequacy substantiation.

Special attention shall be paid to analysis of any failures resulting in occurrence of initiating events as well as analysis of failures resulting in non-performance of the assigned functions by the system (component).

In case the initial calculation information and analysis are related to the personnel's actions the analysis results with regard to safety impact of any erroneous personnel's actions shall be presented.

Information on the instrumentation and hardware installed to prevent or mitigate the consequences of any operational occurrences and accidents shall be provided.

7.2.2. Central control room

The requirements of par. 7.2.2.1 - 7.2.2.5 are similar to the requirements stated in par. 7.2.1.1 - 7.2.1.5.

7.2.2.1. Purpose, configuration and design basis

The requirements are similar to the requirements stated in par. 7.2.1.1.

7.2.2.2. Description

The following shall be provided:

- general layout of the CCR;

- configuration of the CCR consoles (instrumentation and hardware, controls, transducers), panels and boards with the control and monitoring devices installed on them;

- information on location of safety-related I&C systems and information necessary to substantiate ergonomic requirements for their design, arrangement of information and body fields on the control room panels and boards of the control station (stations).

Special attention shall be paid to information on substantiation of the following solutions:

- registration of the control personnel's actions in emergency situations;

- automatic provision of information on the state of process equipment and safety-related automation devices to the operator;

- independent operability check for process equipment and safety-related automation devices performed by the operator in the course of functioning;

- the list of functions performed automatically with submittal of the relevant information to the operator;

- the list of functions performed by the operators. Information substantiating duplication of automatically implemented functions with functions performed with involvement of the operator shall be presented.

It should be demonstrated how control and monitoring of the RP operation and any other NPU systems including the safety systems is arranged from the CCR under normal operation conditions and in case of accidents.

Instrumentation and hardware making the information suitable for the operator to perform any necessary actions aimed to ensure safety shall be described.

The level for solution of any problems related to the human-machine interface shall be substantiated.

Information on substantiation of the workspace sufficiency for all operating personnel both under normal RP operation conditions and in case of any emergency situations shall be provided.

Information on ergonomic and anthropometric arrangement of the workplaces for operators shall be provided.

The following shall be substantiated for the information fields of the operator's workplace:

- location of the display means for safety-related information on the panels and boards of the control station (stations);

- distinctive coloring of the safety-related information display means;

- convenience of the operator's observation over the displayed safety-related information (zones of vision, size of scales, figures and other symbols);

- reliability of the applied lighting for scales, figures and other symbols on the display equipment.

The following shall be substantiated for the body fields of the operator's workplace:

- location of the controls (buttons, switches, etc.) for safety-related actuators on the body fields of the control room panels and boards of the station (stations) with due regard for convenience of observation over the displayed information and distinctive coloring of the controls for safety-related actuators;

- location of the devices for authorized access to the controls of safety-related actuators in case

any such requirements are prescribed.

The following shall be substantiated:

- color, sound, and other distinctive characteristics of alarms that shall be well identified by the operator and have uniform interpretation at all control rooms of the NPU;

- application of the CCTV means;

- application of the information means intended for use by all operators on the watch;

- ergonomics of the technical solutions for manual and automated information recording by the operator at the workplace.

7.2.2.3. Commissioning works

The requirements of par. 7.2.2.3 are similar to the requirements stated in par. 7.2.1.3.

7.2.2.4. Maintenance

The requirements of par. 7.2.2.4 are similar to the requirements stated in par. 7.2.1.4.

Completeness and scope of metrological support for the CCR devices, constituent parts and components shall be substantiated.

7.2.2.5. Safety analysis

The requirements of par. 7.2.2.5 are similar to the requirements stated in par. 7.2.1.5.

Reliability analysis results for all components and constituent parts of the CCR shall be provided, and selection of the parameters to be displayed for the operator under normal operation conditions and in case of any operational occurrences and accidents shall be justified. It shall be demonstrated that the selected and displayed parameters ensure submittal of unambiguous and reliable information on compliance with the NPU safe operation limits and conditions to the operator as well as identification and diagnostics of the SS actuation and functioning.

Analysis results for impact of the CCR habitability and survivability supporting systems on its reliability and operability shall be provided.

Results of the analysis shall be provided to confirm that any common cause failures of the CCR and the ECS are prevented.

Analysis shall be provided to demonstrate that the operator has sufficient information for performance of any manual operations required for safety (for example, optimal location of the controls, manual operations for the safety hardware servicing, potential unexpected post-accident actions and state monitoring for the safety hardware) and sufficient time to make and implement correct decisions.

Information enabling to define that the operator has the possibility to read the data and indications of instruments for monitoring of the parameters in the reactor, the coolant circulation system, the reactor containment and safety assurance process systems in all RP operation modes including expected operational states and emergency modes shall be provided.

7.2.3. Emergency cooldown station

The requirements of par. 7.2.3.1 - 7.2.3.5 are similar to the requirements stated in par. 7.2.2.1 - 7.2.2.5.

7.2.3.1. Purpose, configuration (instrumentation and hardware, controls, etc.), design basis

The requirements of par. 7.2.3.1 are similar to the requirements stated in par. 7.2.2.1.

7.2.3.2. Description

The requirements of par. 7.2.3.2 are similar to the requirements stated in par. 7.2.2.2.

Special attention shall be paid to the information demonstrating that control from the ECS ensures bringing of the reactor into sub-critical state, heat removal and long-term maintenance of this state, actuation of safety systems and obtaining of the information on the reactor state.

Independence of the ECS from the CCR shall be substantiated by description of the adopted arrangements and technical solutions.

The following shall be also provided:

- the ECS structure;

- general layout of the ECS;

- configuration of the ECS panels with the control and monitoring devices installed on them;

- the ECS console (if any is provided);

- the ECS console boards with the control and monitoring devices installed on them (if any are provided);

- impossibility of the RP control from the ECS in case the CCR is operable;

- safety-related functions implemented by the ECS.

The arrangements aimed to prevent impact of the ECS on the RP and the CCR in case of any fire in the ECS shall be demonstrated.

Information on location of the safety-related control and monitoring devices and information required to substantiate the ergonomic requirements for their use shall be provided.

7.2.3.3. Commissioning works

The requirements of par. 7.2.3.3 are similar to the requirements stated in par. 7.2.2.3.

7.2.3.4. Maintenance

The requirements of par. 7.2.3.4 are similar to the requirements stated in par. 7.2.2.4.

Information shall be provided to substantiate solutions on the regulations for the ECS operability maintenance under normal operation conditions.

7.2.3.5. Safety analysis

The requirements of par. 7.2.3.5 are similar to the requirements stated in par. 7.2.2.5.

The list of safety-related functions implemented from the ECS as well as the information required to substantiate impossibility of the common cause failure for the CCR and the ECS and conditions for the CCR operating personnel moving to the ECS in case of impossibility to carry out control from the CCR shall be provided.

Solutions for assurance of the ECS habitability and survivability in case of design basis and beyond design basis accidents shall be analyzed.

7.3. Safety systems

7.3.1. Control safety systems of the NPU

7.3.1.1. Purpose and design basis

The requirements of par. 7.3.1.1 are similar to the requirements stated in par. 7.2.1.1.

7.3.1.2. Description

The requirements of par. 7.3.1.2 are similar to the requirements stated in par. 7.2.1.2.

Description of each CSS shall contain:

- structure of the system;

- functions performed by the system automatically;

- the system operation algorithms;

- configuration (instrumentation and hardware, controls, sensors, transducers, etc.), structure and characteristics of the system channels;

- description of the system operation principle;

- schemes and drawings showing location of the system parts and components.

Multiple channels of the system, its independence from the instrumentation and control system shall be substantiated; it should be demonstrated that combination of normal operation functions for individual components does not affect performance of their safety functions. All system parts and components shall be described.

7.3.1.3. Commissioning works

The requirements of par. 7.3.1.3 are similar to the requirements stated in par. 7.2.1.3.

7.3.1.4. Maintenance

The requirements of par. 7.3.1.4 are similar to the requirements stated in par. 7.2.1.4.

7.3.1.5. Safety analysis

The requirements of par. 7.3.1.5 are similar to the requirements stated in par. 7.2.1.5.

The information shall contain substantiation of the way to ensure compliance of the CSS with the established safety requirements, including the analysis results with regard to:

- the system operation reliability;

- consequences of its failures;

- consequences of the SSS failures.

7.3.2. RP control and protection systems

The requirements of par. 7.3.2 are similar to the requirements stated in par. 7.2.1.

7.3.2.1. Purpose and design basis

The requirements of par. 7.3.2.1 are similar to the requirements stated in par. 7.2.1.1.

7.3.2.2. Description

The requirements of par. 7.3.2.2 are similar to the requirements stated in par. 7.2.1.2.

Constituent parts and components (instrumentation and hardware, controls, sensors, transducers, etc,) of the RP CPS reactor shutdown systems particularly the ones not performing the emergency protection function under normal operation conditions and in case of any operational occurrences and accidents shall be described.

Description of the RP emergency protection system (EP) shall contain:

- the system structure and characteristics of the channels;

- functions performed by the system automatically;

- functions performed by the operator;

- description of the system operation principle.

The presented materials shall contain:

- the lists of initiating signals for the reactor EP activation;

- description of the executive protection signal formation logic for each parameter;

- description of the redundant protection actuation techniques;

- description of the conditions for authorized access to actuation of protections;

- description of redundancy for the channels implementing the protection functions;

- substantiation of the each reactor EP system structure compliance with the diversity principle;

- description of the protection actuators.

All means supporting normal functioning of the reactor EP systems shall be defined and described in the materials. Besides, the following shall be specified for each system:

- the system operation algorithms;

- configuration, structure and characteristics of the system channels;

- information on the system hardware location.

Information shall be provided to substantiate independence and adequacy of power supply systems for the EP systems under normal operation conditions, in case of design basis accidents and considered beyond design basis accidents. Special attention shall be paid to the information on the adopted procedure for detection and elimination of the EP actuation causes as well as sequence of the operating personnel's actions in order to recover the NPU operability after the EP actuation.

Methodologies for calculation, monitoring and control of neutron flux and reactivity, their constituent parts and components shall be described:

- monitoring channels;

- recording devices;

- reactimeters;

- automatic operability check means for the monitoring channels and malfunction warning alarm;

- the methodology for the nuclear core sub-criticality calculation;

- methods for monitoring of the power density irregularity across the nuclear core;

- the method for in-process calculation of the burnout ratio.

The input calculation information on all parameters and characteristics of the system, the system circuits and constituent parts, their layout and arrangement drawings shall be provided.

7.3.2.3. Commissioning works

The requirements of par. 7.3.2.3 are similar to the requirements stated in par. 7.2.1.3.

7.3.2.4. Maintenance

The requirements of par. 7.3.2.4 are similar to the requirements stated in par. 7.2.1.4.

7.3.2.5. Safety analysis

The requirements of par. 7.3.2.5 are similar to the requirements stated in par. 7.2.1.5.

In the course of safety analysis special attention shall be paid to impact of any failures occurring at the ship that can cause loss of the possibility for the reactor shutdown systems and devices to perform their functions, as well as failures resulting in occurrence of initiating events (unauthorized control signals) or failure of the system to perform the assigned functions. In this case it should be demonstrated that the RP CPS can prevent uncontrollable transfer of the reactor into the critical state.

The analysis shall contain review of each system functioning in accordance with the approved lists of initiating events in case of design basis and beyond design basis accidents.

The materials of the section shall contain the analysis results for:

- reliability of the reactor EP system functioning;

- consequences of its failures;

- consequences of the SSS failures (power supply and transformation, ventilation, conditioning, etc.).

The following shall be specified for each system:

- description of diagnostics for the information display channels;

- substantiation of the fact that the operator has sufficient information (for example, position, changes in the position of control devices in the nuclear core, operability of the monitoring channels, safety-related parameters, power recording, etc.) for remote performance of the operations necessary for safety assurance; it should be demonstrated that the concept of the 30-minute operator's non-intervention in case of emergency situations is implemented in the I&C design.

It shall be demonstrated that the design provides for automatic activation of the reactor safety

systems upon occurrence of all events requiring their fast response; the automatically activated systems shall be able to maintain the safe state of the RP within 30 minutes without any intervention of the operator.

It should be demonstrated that control of safety systems by the operator is allowed on the condition that any error of the operator will not have any adverse effect on the SS operation and will not prevent the protection activation.

In this case the way to implement prohibition (automatical or through the use of special-purpose instructions) of remote control inconsistent with the emergency control algorithm shall be demonstrated.

Analysis shall be presented in order to define provision of information on the following parameters to the operator in all RP operation modes:

- parameters determining the reactor state;

- the coolant circulation and heat removal system;

- process safety systems including automation and control means;

- parameters determining the conditions in the reactor containment.

7.4. Monitoring of the radiation situations in the rooms with the NPU equipment

The requirements of par. 7.4.1.1 - 7.4.1.5 are similar to the requirements stated in par. 7.2.1.1 - 7.2.1.5.

7.5. Communication and annunciation systems

7.5.1. Purpose and design basis

The requirements of par. 7.5.1 are similar to the requirements stated in par. 7.2.1.1.

7.5.2. Description

The requirements of par. 7.5.2 are similar to the requirements stated in par. 7.2.1.2.

Description of the systems and means for preventive and emergency warning of the NPU personnel and the ship crew shall also contain:

- the list of warning signals with indication of any light, sound and other effects accompanying them in order to attract attention of the personnel;

- technical characteristics of the means intended to attract attention of the personnel (flashing frequency, color, pitch of tone, etc.).

Information on the adopted preventive and emergency warning system for the personnel shall contain the rules for usage of the warning signal system in emergency situations.

Information on the communication means and annunciation systems, particularly redundant ones, intended for the ship NPU control arrangement under normal operation conditions, in case of any design basis and beyond design basis accidents shall be provided.

It should be demonstrated that the CCR communication with the pilot room, the ECS room, the

main engine room, the rooms of the main, redundant and emergency generators, the attended rooms of the controlled access area is ensured in case of complete loss of power supply on the ship.

7.5.3. Commissioning works

The requirements of par. 7.5.3 are similar to the requirements stated in par. 7.2.1.3.

7.5.4. Maintenance

The requirements of par. 7.5.4 are similar to the requirements stated in par. 7.2.1.4.

7.5.5. Safety analysis

The requirements of par. 7.5.5 are similar to the requirements stated in par. 7.2.1.5.

7.6. Non-safety-related systems

7.6.1. Description

The following information shall be provided for non-safety-related control and monitoring systems and means:

the list of these systems and means, their purpose, technical characteristics.

7.6.2. Safety analysis

It should be demonstrated that these systems are not required for safety assurance.

8. POWER SUPPLY

Information confirming functional development and reliability of the supporting power supply systems, sufficient capacity, availability of multiple channels, independence, stability under external and internal impacts, the possibility to perform maintenance, testing and repair, compliance with the requirements of safety standards and rules based on the analysis of the power supply system functioning with due regard for human errors as well as under normal operation conditions, in case of any operational occurrences, design basis and beyond design basis accidents shall be provided in the SAR. Qualitative and quantitative reliability analysis for the power supply systems shall be presented.

The basic principles for design and operation of the electric systems shall be specified, their list and schemes shall be provided.

Deviations from the requirements of standards and rules, the reasons for these deviations and the compensatory measures shall be specified for each system.

Completeness of the descriptions, presented technical data and calculations shall be sufficient to perform independent review of the electrical and technical part of the NPU-powered ship design.

8.1. Main electrical system

It should be demonstrated that activation of the reactor EP and loss of the ship maneuvering ability is prevented in case of failure of any components of the main generator, its motor and the associated mechanisms, as well as in case of failure of any component in the system switchgear.

The information on the system shall be presented in accordance with par. 1.4.1 - 1.4.5.

8.1.1. Main electrical system diagram

It is necessary:

- to provide the main electrical system diagram with its division into separate parts and substantiate that any operational occurrences and design basis accidents in any part of the system will result in automatic switch-off of this part, and it will not affect normal operation of any other parts;

- to demonstrate that the mechanisms and systems of the operating NPU are supplied from two parts of the main electrical system, and their switching from the main power supply to backup one is performed automatically and does not cause any deviations of the system parameters or activation of the preventive warning; it should be also demonstrated how the backup generator takes the load from any part of the main electrical system in case of necessity, how parallel operation of the backup generator and the main generators is arranged in the course of load transfer without any supply voltage and frequency drop for the consumers below the established limits;

- to provide power supply diagrams for the NPU consumers as well as for any other redundant essential consumers;

- to demonstrate location of the main and backup generators, the main switchboards on the ship, physical separation of the switchgear rooms, power supply sources and cable routes for the multi-channel power supply system, their protection against external impacts.

8.1.2. Characteristics of the system

It is necessary:

- to specify the capacity of the main electrical system components and to demonstrate that the total capacity of the operating main generators of each system part is sufficient for complete power supply of all consumers required to maintain normal operable state of the ship. All planned EPS operation modes from the pier-side moorage with the NPU in the off state and power supply from the onshore sources to the NPU shutdown and switching to power supply from the onshore power supply sources as well as the optimum possible EPS operation modes in case of failure of any power supply source including power supply from onshore sources (other ship) and power supply to the shore (other ship) shall be described;

- to demonstrate compliance with the requirements of standards and rules stipulating that in case of any voltage loss on the buses of any main switchboard the time for start-up and load taking by the backup generator shall ensure safe NPU operation, and the total capacity of the backup generator and the main generator remaining in operation shall be sufficient to ensure power supply of the consumers required to maintain normal operable state of the ship. Power consumers disconnected in this case shall be listed, and it should be demonstrated that their disconnection is insignificant for safety of the operating NPU and the ship;

- to demonstrate that the backup generator capacity is sufficient for shutdown, cooldown, elimination of the operational occurrence or design basis accident in the RP and its subsequent activation.

8.1.3. Fire safety of the main EPS equipment

Impact of any fire hazard of the main electrical system equipment on the ship safety shall be analyzed.

8.1.4. Controls of the main EPS

It should be demonstrated that any damage will not cause inoperability of the control and

monitoring for more than one part of the main electrical system.

8.1.5. Possibility for testing and maintenance

The following information shall be presented in the subsection:

- continuous automatic diagnostic self-control of the EPS and its components;

- frequency of testing, testing methods and programs, controlled parameters, alarm activation setpoint values;

- the possibility to perform testing on operating or shut-down equipment;

- types and time limits for maintenance of the switching equipment, protection and automation cables;

- operability recovery techniques;

- time limits for replacement of the equipment and cables with exhausted lifetime;

- accessibility for maintenance and testing with due regard for the radiation hazard conditions.

8.1.6. Safety analysis

Descriptions of the calculation programs, the input data used to calculate the algorithms of the EPS automatic and remote control, calculation results, as well as information on validation and verification of the programs shall be provided in the SAR.

Results of any experimental tests shall be interpreted and compared with the calculation analysis.

Functioning of the EPS under normal operation conditions, in case of any operational occurrences and design basis accidents, switching to operation from the emergency power network and other associated power networks with due regard for their possible failures shall be described.

The list and analysis of design failures of the components including human errors shall be provided, and impact of the failure consequences (including common cause failures) on the EPS operability and the NPU safety shall be assessed.

8.2. Emergency electrical system

Technical data of the system and its components shall be presented in this subsection; it is also necessary:

- to provide the emergency electrical system diagram with generators independent from the RP. It should be demonstrated that the emergency electrical system performs safety functions with due regard for the single failure principle under all design conditions;

- to demonstrate that the emergency electrical system capacity if sufficient for the RP shutdown and cooldown as well as its maintenance in the safe state for 30 days;

- to provide electrical and technical characteristics of each consumer supplied from the emergency electrical system with indication of the required period of its operation without any power supply from the main electrical system;

- to provide the list of auxiliary consumers requiring power supply from the emergency electrical system in case of any loss of power supply from normal operation sources with indication of the

permissible power supply characteristics for each consumer:

a) duration of power supply interruption;

b) maximum (minimum) permissible current voltage and frequency variation with indication of the permissible duration;

c) current waveform changes and their duration;

d) the period upon expiry of which voltage may be re-supplied to the consumer and any other restrictions (if any) from the I&C and CSS systems;

The nameplate data of each consumer shall be provided with indication of the period of its operation without any main and (or) backup power supply from normal operation sources:

- requirements for fire safety, fire and explosion protection of the equipment and hardware and fire resistance of the emergency power supply system structures and the SS electrical equipment shall be specified;

- to specify operating conditions for the electrical equipment, switching devices, SS cables and the emergency electrical system in normal and emergency NPU operation modes with regard to temperature, humidity, pressure, radiation and other external impacts with indication of the exposure time;

- to demonstrate that each emergency generator is started automatically upon loss of voltage on the relevant bus of the group switchboard and by the reactor EP reset signal;

- to confirm independence of the emergency power distribution systems (from at least two emergency generators), automatic start-up of the generators by the voltage or frequency drop signal and (or) the reactor EP reset signal; in this case no direct synchronization of the power sources is required;

- to demonstrate that the emergency electrical system can be activated from the CCR and the local emergency generator control boards;

- to confirm that the emergency electrical system will take the load within the period of time determined by the reactor safety conditions, and in this case no direct synchronization of the power sources is required in emergency conditions;

- to provide information on fire safety assurance including description of the automatic fire detection and extinguishing systems with the results of the relevant calculations;

- to substantiate the continuous operation duration for any sources with limitations imposed by fuel stocks.

8.3. Intermediate power sources

Technical data of the system and its components shall be presented in this subsection; it is also necessary:

- to provide description of the connection diagram for the intermediate power sources and the list of instruments for measurement of the RP parameters, radiological control and any other safety-related measuring and indication devices supplied from these sources within 30 minutes;

- to indicate location of the intermediate power sources on the ship and to confirm their operability under normal operation conditions, in case of any operational occurrences and accidents. In

case any storage batteries are used as the intermediate power sources the charging device scheme shall be provided, and compliance of its characteristics with the RD requirements shall be demonstrated.

In case no intermediate power sources are provided on the ship it should be demonstrated that the SRS consumers have continuous power supply from other electrical systems with due regard for the single failure principle in case of any operational occurrences and design basis accidents.

8.4. Cable network of the controlled access area

The following information shall be presented in the subsection:

- it should be demonstrated that the type of electric cables, their laying techniques as well as sealed cable penetrations through the containment, shielding and other partitions of the controlled access area comply with the requirements of standards and rules;

- it should be confirmed by testing that cable penetrations through the containment and shielding do not reduce their tightness. The possibility for local leak-tightness monitoring for the penetrations in the course of testing and operation shall be provided;

- the ambient parameters in case of design basis accidents shall be specified; it should be demonstrated that cables retain their operability under these conditions;

- the cable routing diagrams for the controlled access area shall be provided in order to confirm that the cables to the redundant consumers and safety systems are laid separately from the main power routes.

8.5. Lighting

Technical characteristics of the lighting systems and their components shall be specified in this subsection, implemenetation of safety principles shall be demonstrated, and the following shall be also provided:

- schemes and descriptions of the main and emergency lighting of the controlled access area rooms as well as emergency lighting of the CCR, the ECS, etc., power supply diagrams for lighting fixtures, location of switchboards, switches and circuit breakers;

- compliance with the requirements of standards and rules shall be demonstrated.

8.6. Connection to the external grid

8.6.1. Power supply from the external power source

The scheme of the main ship electrical system power supply from the external source shall be presented, and location of the relevant connections, boards and switching devices shall be demonstrated.

8.6.2. Power supply to the external consumer (the onshore grid, the power system of other ship)

In case power supply to any external electrical system (a damaged ship, a floating power unit, an onshore consumer) is provided sufficiency of the controlled capacity of the main ship electrical power system, the possibility to limit the power demand in the external power network, the minimum permissible insulation resistance values for these networks shall be specified. Besides, the cases when the necessity to limit the ship NPU power may occur in the grid as well as the rate and duration thereof

shall be specified.

The possibility for automatic or manual isolation of the ship electrical system from the external grid with switching to the auxiliary power supply mode shall be demonstrated.

Impact of any potential types of external consumer failures on the NPU safety shall be analyzed.

9. AUXILIARY SYSTEMS OF THE SHIP NPU

9.1. Set of refueling equipment for nuclear cores

The set of refueling equipment is not a regular NPU system but the refueling works for the reactor nuclear cores are performed in the ship rooms and represent a range of specific potential nuclear and radiation hazardous works; their safe performance depends on may factors determined by the refueling equipment.

The particular configuration of the set shall be specified in the introduction to the section, including:

a) the nuclear core refueling system;

b) the SNF handling system consisting of:

- the SNF storage system in the repository located at the nuclear maintenance ship or in the building specially constructed on the shore for this purpose (SNFS) (if any);

- the set of refueling equipment, the devices for the nuclear core refueling, NF storage and transportation;

c) arrangement of the NF accounting at all stages of handling in the course of nuclear core refueling.

9.1.1. Nuclear core refueling equipment

Requirements established for the refueling equipment, special-purpose appliances, tools, measuring means, instruments and devices ensuring performance of the refueling operations shall be specified.

9.1.1.1. Description of the refueling flow diagram and mechanical aids

Restrictions with regard to the meteorological conditions for refueling as well as technical conditions of the RP (cooldown to the heat emission levels accommodated by the repair cooldown system is completed, temperature and pressure in the primary circuit, the water level in the pressurizers of the sealed circuit and (or) in the reactor vessel in case of unsealed circuit are controlled) shall be specified.

The process flow diagram of the refueling operations, equipment, devices and components shall be described.

Comparative analysis of the process flow diagram for the refueling operations with similar domestic or foreign projects shall be provided.

Priorities in solution of safety issues shall be demonstrated in comparison with other similar ship reactor refueling projects.

The design process flow diagram of the refueling operations shall be described with indication of

differences (if any) in the refueling process and special-purpose equipment applied in each particular case.

In this case the following shall be specified:

- the selected refueling technique (channel-by-channel, with shielding assembly, dry or wet) and its substantiation;

- state of the containment or its sealing in the course of refueling (regular and (or) maintenance covers, airlocks, doors);

- the design refueling frequency, scope and procedure;

- the actual reasons for the planned nuclear core refueling and their substantiation;

- engineering features provided in the design for biological protection, prevention of any accidental ingress of foreign objects into the reactor and the primary circuit in the course of refueling and repair works;

- configuration of the refueling equipment with its adequacy substantiation and also with indication of the requirements for this equipment ensuring safe handling of fuel assemblies, particularly in case of any failures and damages.

The following shall be demonstrated:

- measures aimed to prevent formation of critical mass at all process stages of the nuclear core refueling with both irradiated and non-irradiated fuel assemblies;

- the sub-critical neutron flux monitoring means with at least two measuring channels at all refueling stages;

- measures aimed to prevent any deformations, destruction or falling of fuel assemblies;

- engineering features aimed to prevent falling of fuel assemblies in case of any power supply loss;

- measures to prevent application of impermissible forces to the fuel assemblies in the course of their withdrawal or installation;

- peculiarities of refueling in case of reaching (exceedance) of the safe operation limits for the unloaded nuclear core. Operability of the refueling equipment components shall be substantiated, and information on the systems related to functioning of the nuclear core refueling set shall be specified.

Information on the following systems shall be presented in accordance with par. 1.4.1-1.4.5:

- closed circuit television for the refueling monitoring with the list of refueling operations controlled through the use of closed circuit TV;

- monitoring of water level and temperature in the reactor;

- monitoring of fuel element claddings (incoming control);

- power supply, particularly emergency one;

- operating and emergency lighting;

- fire extinguishing;

- ventilation and air purification;

- communication, annunciation and alarms;

- decontamination (if any).

9.1.1.2. NF accounting arrangement

The NF accounting system shall be described, and its compliance with the requirements of standards and rules shall be demonstrated.

9.1.1.3. The follwoing requirements shall be specified:

- for the qualification and training level of the refueling personnel;

- for the supporting systems.

9.1.1.4. The list of nuclear-hazardous works and technical requirements for their performance shall be provided.

9.1.1.5. Tests and inspections

The basic requirements related to quality assurance for the system and its components in the course of manufacturing, construction and installation shall be provided.

The scope and methods of incoming control, inter-departmental, commissioning tests, their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation used for the purpose of testing shall be presented and substantiated.

The results of analysis for compliance with the requirements, principles and criteria of the relevant safety regulatory documents in the course of nuclear core refueling shall be provided.

9.2. Systems supporting the NPU operation

The list of ship systems supporting the NPU operation including the power supply, air systems, ventilation, conditioning, physical protection, explosion protection, controlled RSb discharge, drainage and emergency drainage of the rooms shall be provided; schemes of branches from the ship systems shall be presented, and their redundancy and compliance with the RD requirements for reliability assurance shall be demonstrated.

The supporting systems shall be described in accordance with the standard description structure given in Subsection 12.4.

10. RADIOACTIVE WASTE MANAGEMENT

The following information shall be presented in the section:

- safe and reliable handling of all types of radioactive wastes generated in the course of the NPU operation including design basis accidents;

- the paths of their supply to the rooms and the environment; in this case it should be specified that the RW quantity measurement methods and special-purpose means for reduction of leakage comply with the requirements of standards and rules;

- techniques for storage, processing and handling of radioactive wastes;

- RW generation sources (fission products, induced radioactivity in liquids and metals, consumables and clothes), amounts and physical and chemical composition of SRW, LRW and GRW;

- planned annual amounts of generated SRW, LRW and GRW and their aggregate activity by groups;

- RW segregation and sorting methods;

- selection of the RW handling system;

- RW storage facilities;

- RW conditioning and quality control methods;

- control methods for chemical and radionuclide composition of radioactive wastes;

- reliability substantiation for the protective barriers;

- possibility of sampling at all RW handling stages;

- explosion and fire safety assurance in the course of RW handling;

- radiological control in the course of RW handling;

- possibility for the equipment decontamination;

- monitoring of the RSb ingress into the environment;

- availability of the alarm, interlocking and protection systems in the RW storage facilities;

- lifting and handling equipment;

- availability of special-purpose containers for radioactive wastes;

- the labeling system for the RW packages.

Non-wastefulness level of the applied technologies shall be assessed from the viewpoint of RW generation.

10.1. SRW management system

It is necessary:

- to describe all systems representing potential sources of SRW generation in the course of normal operation, in case of any operational occurrences and design basis accidents;

- the main safety principles and criteria implemented in the design, barriers preventing ingress of radioactive substances into the environment in excess of the limits established by the federal rules and regulations in the area of atomic energy use shall be specified;

- it should be substantiated that the design of SRW handling systems provides for the engineering features ensuring dosimetric and process control of the SRW state and content in accordance with the requirements of standards and rules, the storage facilities for non-standard SRW and special-purpose rooms for SRW collection and segregation in accordance with their classification are available, special-purpose containers and lifting and handling equipment are used, and the climatic conditions on the ship prevent any damage of packages and changes of the SRW form, the system of administrative and technical measures prevents any unauthorized access to SRW.

10.2. LRW management system

The capabilities of the ship systems for LRW handling in the course of normal operation and in case of any operational occurrences and design basis accidents shall be described; it is also necessary:

- to present the main safety principles and criteria implemented in the design, barriers preventing ingress of radioactive substances into the environment in excess of the limits established by the federal rules and regulations in the area of atomic energy use;

- to demonstrate that design of the tanks enables to accommodate all LRW amounts provided in the NPU design under normal operation conditions and in case of design basis accidents, each tank is equipped with the devices for LRW receiving and complete emptying, process control (temperature, pressure, level in the tank, the upper position alarm) including leakage monitoring, the hydrogen monitoring means, fire extinguishing means; the LRW handling system complies with the requirements of standards and rules, and the design provides for LRW state monitoring at all stages of their handling;

- to demonstrate that the LRW handling system ensures non-exceedance of the permissible surface contamination levels for the ship rooms and radiation in all operation modes and in case of any design basis accidents at the NPU;

- to describe the measures provided in the design to control the LRW ingress and to substantiate adequacy of the preventive measures;

- to substantiate safety assurance in operation of the LRW handling systems under low temperature conditions and peculiarities of safety assurance in the course of LRW transfer to onshore and floating facilities for further processing.

10.3. GRW management system

It is necessary:

- to specify all ship systems representing potential sources of RSb releases into the environment in the form of gases or aerosols including ventilation systems of the controlled access area and the supervised area;

- to substantiate the ship capabilities for GRW handling in all NPU operation modes including emergency situations;

- to specify the main safety principles and criteria implemented on the ship, safety class and category in accordance with the adopted safety classification;

- to indicate the barriers preventing any ingress of radioactive substances into the environment above the limits established by the standards and rules;

- to substantiate the capability of the GRW handling systems to ensure safe GRW handling within the entire NPU service life and their compliance with the standards and rules;

- to provide assessment confirming that the systems have sufficient capacity and the required redundancy to ensure air purification from radioactive substances in all cases with loss of fuel element integrity or failures of safety systems;

- to demonstrate that the GRW handling system ensures non-exceedance of the permissible release limits in all operation modes and in case of any design basis accidents at the NPU;

- to describe the measures provided in the design to control the GRW supply from the process equipment and to substantiate adequacy of the preventive measures;

- to present the list of equipment and systems where generation of explosive gas concentrations is possible as well as the design pressure and substantiation of the equipment adopted in the design;

- to describe process instrumentation and hardware;

- to assess compliance with the RD requirements, safety principles and criteria.

11. RADIATION SAFETY

Information on the ways to ensure radiation safety shall be provided in this section:

- in case of external exposure from penetrating radiation - gamma quanta and neutrons; the sources of such exposure include the nuclear core, the structural materials of the reactor and the equipment containing radioactive substances;

- in case of internal exposure for the human organism (oral and inhalation ingress of radioactive substances).

It is necessary:

- to demonstrate that the individual exposure limits for the personnel in all normal operation modes and in case of design basis accidents will not exceed the established limits and the collective annual exposure dose, that they are reduced to the minimum reasonably achievable level for both the crew and the public, and ingress of radioactive substances into the environment will not result in exceedance of any exposure quotas for the public recommended by the regulatory documents;

- to present the programs for the radiation situation monitoring, the individual dosimetric monitoring program and the environmental radiological control program;

- to specify quantitative values of the radiological criteria used to identify occurrence of any emergency situation or accident.

Conclusions on compliance of the implemented measures with the RD requirements for radiation safety shall be made in each subsection.

11.1. Achievement of the lowest reasonably achievable level of occupational exposure (ALARA principle)

11.1.1. Radiation safety concept

It is necessary:

- to describe principles, criteria, calculation methods, engineering features and administrative arrangements used to ensure protection of the personnel, the public and the environment against impermissible radiation exposure;

- to demonstrate that compliance and application of the engineering features and administrative arrangements is justified by operation experience for similar NPUs and will not result in exceedance of any exposure levels regulated in the RD, and the impact of hazardous factors in all ship NPU operation modes and in case of accidents will remain at the lowest reasonably achievable levels. Achievable exposure levels shall be presented in the form of the collective annual dose for all personnel and the public and the doses for individual categories of the personnel under normal operation conditions and in

case of design basis accidents;

- to demonstrate efficiency of the protection systems and their sufficiency to guarantee insignificant damage to the personnel health, the public and the environment.

In this case the following limits shall be specified:

- individual exposure dose for the personnel;

- collective annual exposure dose for the personnel;

- public exposure.

11.1.2. Design basis

It is necessary:

- to describe the ways to reduce the radiation exposure down to the lowest reasonably achievable level through the use of the developed radiation protection principles, selection of the technical and administrative solutions applied in the design of the RP and ship components;

- to demonstrate how the accumulated experience in designing and operation of other NPUs is applied in the design to reduce the radiation exposure levels down to the lowest possible values with indication of specific arrangements and substantiation of their feasibility in the design;

- to substantiate efficiency of any solutions provided in the design in order to reduce exposure levels in the controlled access area rooms and to reduce the time of the operating personnel stay in these rooms, particularly to decrease the number of RSb sources, to enhance the radiation protection, to reduce the time of maintenance, to facilitate access to the equipment, to simplify operational procedures and also to reduce and simplify any other actions required within the operation period;

- to specify the ship zoning criteria, to provide the list and brief characteristics of the rooms referred to the controlled access area;

- to present the list of special-purpose technical solutions aimed to ensure compliance with the requirements of NRB-99 for the exposure dose limits in case of beyond design basis accidents.

11.1.3. Arrangement of operation

The following shall be demonstrated in the subsection:

- the ways to consider the requirements for non-exceedance of the occupational exposure doses defined in the regulatory documents in arrangement of the NPU operation, as well as to consider the requirements for arrangement of operation of any other similar NPUs in development of equipment, biological protection and design of the NPU;

- radiation criteria used in the development of manuals and engineering features for performance of radiation-hazardous works, including maintenance, in-service inspections, metal state control, refueling of the reactor nuclear cores, works with radioactive wastes in order to reduce the occupational exposure doses;

- the ways to limit internal and external exposure for the ship personnel and to arrange separation of workplaces and rooms in accordance with the zoning principle;

- the list and quantitative values of the operational criteria such as total specific activity of fission products in the primary circuit coolant, specific volumetric activity of air in periodically attended rooms, contamination levels for the surfaces of the rooms and equipment installed in these rooms;

- the list and quantitative values of the NPU operation process parameters (the operation limit with regard to fuel element damage, the coolant leakage value, etc.) to be maintained in order to ensure the lowest reasonably achievable levels of radiation exposure.

11.2. Radiation sources

11.2.1. Equipment containing radioactive substances

Information on the content of radioactive substances in the equipment components (except for the RW handling systems described in Section 10) representing sources of radiation taken into account in the biological protection calculations and design shall be provided.

The following shall be described:

- the reactor nuclear core as the source determining the ionizing radiation levels in the course of the reactor power operation in the rooms behind the biological protection where any operating personnel can be present as well as the source of fission products supplied to the primary circuit;

- materials of the pressure vessel and other structural components of the reactor as the source of capturing and activation gamma radiation;

- the primary circuit as the source of activation products for the primary circuit coolant and activated corrosion products as well as fission products entering the coolant due to any defects of fuel element claddings;

- the secondary circuit as the source of radioactive substances in case of any circuit-to-circuit primary circuit coolant leakages;

- the third circuit as the potential source of radioactive substances in case of any operational occurrences and design basis accidents;

- other NPU systems and equipment capable of containing or accumulating radioactive substances in the course of operation;

- other radiation sources including start-up neutron sources, sources for calibration of instruments and devices, sources for gamma-radiography, nuclear reaction by-products and any other sources requiring radiation protection.

Description of the radiation sources (except for the reactor nuclear core) shall contain the table of radionuclide composition and radiation energies, information on the activity, geometric parameters of the source as well as the input data for determination of the specified values.

Information on the radionuclide composition, amount and physical and chemical forms of all ionizing radiation sources generating the annual effective dose of more than 10 μSv and the collective annual dose of more than 1 man-Sv under any conditions of their use shall be provided in the SAR.

Correspondence of the fission product ingress into the coolant to the rated operation limit with regard to fuel element damage in the course of power operation shall be substantiated. Increase of the fission product ingress to the coolant from the nuclear fuel in case of any emergency situations and transient modes shall be taken into account.

Information shall be presented in such a way so that it could be used as the input data for the biological protection calculations.

Location of all radiation sources as well as any possible and actual RSb migration paths shall be

indicated on the general layout drawings and plans of the power unit.

11.2.2. Sources of gaseous radioactive substances

Sources of any releases of gaseous radioactive substances into the atmosphere of the controlled access area rooms taken into account in development of the protective measures and assessment of occupational exposure doses shall be described. Apart from the sources existing under normal operation conditions, the sources appearing due to failures of the main equipment and in the course of repair works (opening of the reactor, SNF transportation, etc. and also in the course of the ship NPU decommissioning) shall be described.

The description shall contain calculated concentrations of radioactive gases and aerosols expected in normal operation and transient modes, in case of any anticipated operational occurrences in the controlled access area and in the course of the ship NPU decommissioning.

Models, parameters and input data required to calculate concentration of radioactive gases and aerosols shall be provided. In case of input data absence operational data on similar NPUs may be used.

11.3. Design peculiarities with regard to radiation protection

11.3.1. Equipment layout and arrangement of the ship rooms

The plan (on scale) of the ship rooms with arrangement of the process equipment representing the source of radiation as well as all radiation sources described in Section 10 and Subsection 11.2 shall be provided.

The following shall be indicated on the plan:

- boundaries of the controlled access area and division of its rooms into non-attended, periodically attended and attended as well as rooms of the supervised area;

- location of airlocks, medical stations, etc.;

- schemes of the personnel and transport movement, delivery of clean equipment and materials and removal of contaminated ones;

- location of the storage facilities for contaminated equipment, decontamination areas, places for SRW collection;

- location of the radiological control system sensors;

- location of laboratories (if any) for analysis of radioactive media samples (chemical, radiochemical, radiometric, spectrometric, if any), the individual dosimetric monitoring laboratory, the repair and calibration laboratory (workshop, if any), the storage facilities for ionizing radiation sources. Classification of the ship zones and rooms adopted in the design and used as the basis for design of the biological protection against penetrating radiation and prevention of radioactive contamination of the air in the attended rooms of the controlled access area shall be presented.

The rooms of the controlled access area shall be arranged in such a way so that the personnel would not have to pass through any zones with high dose rates in order to gain access to the zones with low dose rates and through the zones with high contamination level - to get into the zones with lower contamination level.

11.3.2. Structural peculiarities of the systems and equipment components

Design characteristics of the equipment and components enabling to reduce the exposure doses

in accordance with ALARA principle shall be provided, and impact of these characteristics on the basic requirements for the operation procedure described in par. 11.1.3 shall be illustrated through examples.

The description shall include the sections demonstrating structural peculiarities that reduce maintenance or other operations in radiation fields, decrease intensity of the sources and also provide quick entrance to the rooms, easy access to the workplace, remote performance of the operations, reduction of the personnel stay time or any other measures aimed to reduce exposure for the personnel.

Description of the methods used in the design in order to reduce generation, distribution and accumulation of the activated corrosion products, particularly by reduction of the corrosion and erosion rate for the circuit materials, application of the materials with the minimum possible cobalt content in the primary circuit, compliance with the optimal coolant chemistry regimes, minimizing of the dead areas where activation products can accumulate shall be provided. Illustrative examples shall be provided including drawings of the equipment and schemes of pipelines for the components requiring access of the personnel in the course of the NPU power operation. Location of the sampling points, instrumentation and hardware, control panels shall be indicated.

The means and methods aimed to ensure leak-tightness of the equipment and detection of any radioactive media leakage shall be provided, and the list of these means shall be presented.

11.3.3. Biological protection

The following shall be specified:

- information on the biological protection for each radiation source described in Section 10 and Subsection 11.2, including characteristics of the protective materials, thickness of coatings, methods for determination of the protection parameters, geometrical parameters of the source and protection;

- special-purpose protective devices and equipment including containers, casings, screens, loading equipment, etc. used for handling of any radioactive substances;

- calculation programs with the adopted assumptions and the technique used for calculation of the protection, the calculation results, including the design radiation level in the supervised area and the rooms of the controlled access area referred to Category III in the course of normal operation, in case of design basis accidents and during the NPU decommissioning.

11.3.4. Ventilation, purification and conditioning systems

11.3.4.1. It is necessary:

- to describe the main parameters for the design of the controlled access area ventilation systems including the repair ventilation from the viewpoint of the personnel protection as well as any other personnel safety assurance components related to the ventilation systems;

- to demonstrate that the design is based on the principle of separate ventilation for the rooms of the controlled access area and the supervised area in order to prevent ingress of air from the controlled access area to any other areas;

- to specify the measures provided in the design in order to purify the air from radioactive gases and aerosols including the plan of the rooms where purification is carried out and the purification units are installed, as well as the ventilation system schemes;

- to provide the maintenance conditions, to describe the means for control, testing and isolation of the systems, the means for determination of air purification efficiency, replacement and transportation of the spent filter cassettes. Characteristics of the applied air purification means, the

criteria for replacement of the filtering elements, the list of regulatory documents shall be provided. The purification coefficients used in the radiation safety analysis shall be specified; due to dependence of these coefficients from the filtration conditions they shall be assumed for the radiation situation assessment based on the most severe operation conditions for the purification systems (design sizes of the aerosol particles shall be assumed to be equal to the size of the most penetrating particles for each filter; the most unfavorable temperature and humidity conditions out of all possible variants shall be assumed for iodine filters and gas sorbents).

11.3.4.2. The scheme of the service water pipelines for the conditioning system with the location of all coolers in the NPU rooms and technical characteristics of the cooling units shall be provided, their redundancy and operation modes (particularly in case of any design basis accidents including total blackout of the NPU) shall be demonstrated. Calculations of the maximum possible heat emissions from the surfaces of the equipment and pipelines, temperature distribution nomograms for the NPU rooms in the absence of service water circulation shall be provided:

- in the conditioning and cooling water systems;

- for the cooling outer water.

11.3.5. Radiological dosimetric control system on the ship Sampling of radioactive process media

Criteria for selection of radiological monitoring hardware, development of the sampling point scheme and location of the equipment shall be specified. The engineering features provided in the design for radiological control shall be described, including the following hardware:

- continuous monitoring devices based on stationary automated systems and stationary instruments;

- in-process control devices based on portable and mobile instruments and units;

- laboratory analysis devices based on laboratory units, means for sampling and preparation of radioactive samples for analyses;

- individual monitoring of the personnel.

The list of radiological control objects and classification of the control types shall be provided, and the fire resistance category of the system and its equipment components as well as the system category according to its purpose shall be specified.

Description shall include the main technical characteristics (controlled parameters, types and number of sensors, the measurement range, the basic error), information on the metrological support methods and means, information on the alarm units, recording devices and location of sensors, indicating (reading) and signaling devices. The schemes of sampling lines with valves shall be presented.

The design option of the hardware, its compliance with the requirements of fire, electrical and mechanical safety in accordance with GOST 12.1.004-85, GOST 12.2.007.0-75 and GOST 25861-83 shall be demonstrated.

Location of the air sampling points for control of gas and aerosol activity shall be indicated, the air sampling system shall be described, criteria and methods used to obtain representative results of the radioactive gas and aerosol concentration measurement shall be provided.

The capacities of the radiological control hardware with regard to measurement of the radiation situation parameters, particularly intensive radiation and the personnel exposure doses in case of a radiation accident shall be described, the necessity for additional instrumentation for these measurements shall be substantiated, and information on the possibility to ensure redundancy of the

measurement channels, information display and recording means shall be provided.

The list of equipment aimed to monitor contamination of skin, clothes, equipment and various surfaces with alpha-active substances shall be provided; arrangement of the radiation parameters monitoring in the course of SNF unloading and loading of non-irradiated NF shall be also demonstrated, and the list of controlled radiation parameters shall be presented.

Software and mathematical support for processing and presentation of information, the programs ensuring prediction of the radiological consequences of any events on the ship, collection, storage and systematization of data on radiation pollution of the environment and exposure doses for the personnel and the public shall be described.

11.4. Assessment of dose commitments under normal operation conditions and in case of accidents

Annual duration of the personnel stay in the rooms of the controlled access area in the course of normal operation and repair works shall be assessed.

Duration of the personnel stay (in man-hours) and the value of RSb ingress into the human organism due to inhalation shall be assessed for the controlled access area rooms described in Subsection 11.2.2 where gas and aerosol activity is expected.

The annual individual dose (total and separate for external and internal exposure) and dose commitments of the personnel (collective dose) in the course of such functions as operation, maintenance, inspection and examination of weld joints, RW handling, refueling of the reactor nuclear core, repair works shall be assessed in accordance with the regulatory documents.

It should be demonstrated that the exposure doses and dose commitments for the personnel are assessed in real time depending on the ship NPU operation period.

The input data, calculation methods and models and assumptions applied to determine the above-mentioned values shall be specified. In case the estimated (predicted) exposure doses and dose commitments are unacceptably high the arrangements provided in the design for their reduction to acceptable values shall be described.

Information on the exposure doses and dose commitments for the personnel obtained during operation of similar NPUs may be used to assess doses and dose commitments in the course of any unpredictable operations with due regard for certain conservative assumptions.

The annual dose on the controlled access area boundaries at the locations of the main radioactivity sources (the RW storage facilities, points of radioactive releases and discharges) shall be assessed.

Exposure doses for the personnel in case of any design basis accidents shall be assessed. The input data, calculation methods and models and the adopted assumptions shall be specified.

12. SAFETY SYSTEMS

The section shall contain the information on the safety systems provided in the ship NPU design for functioning in case of any operational occurrences and accidents (reactivity-induced, related to malfunctions of heat removal or loss of the primary circuit integrity) and intended for reliable shutdown of the reactor, cooldown of the unit (residual heat removal) as well as for reliable confinement of radioactive releases or the molten corium in case of necessity.

Safety systems shall be activated upon exceedance of the actuation setpoints (prior to exceedance

of the safe operation limits) and in case of any failure of normal operation systems.

Detailed information on the ship NPU safety systems (PSSs, LSSs and SSSs) and their safety functions for prevention of accidents or mitigation of their consequences shall be provided (the requirements for the CSSs are given in Subsection 7.3).

The list and analysis of the functioning of normal operation systems performing any SS functions during accidents shall be presented in this section.

Analysis of emergency situations, design basis and beyond design basis accidents and their consequences shall be provided in Section 15 where the assessment of the SS capability to perform the assigned functions both under normal operation conditions and in case of any failures shall be presented together with the required substantiating materials.

Any assessments included into the SAR shall be reliably correct and complete enough, and all necessary analyses shall be performed. References to the analyses included into any other sections shall be provided if they are related to safety systems.

12.1. Protective safety systems

Information on the protective safety systems for emergency shutdown of the reactor shall be presented in this subsection, including:

1. Emergency nuclear core cooling system

2. Overpressure protection system of the primary circuit

3. Overpressure protection system of the secondary circuit

4. Liquid poison injection system (if any)

5. System for emergency water supply to the reactor

In case there are any other PSSs in the RP they shall be described in accordance with the requirements of the SAR section "General provisions".

PSS description structure:

12.1.1. Design basis

12.1.1.1. Purpose and functions

The following shall be provided:

- purpose of the system with indication of the performed functions;

- additional safety and normal operation functions performed by the system components;

- safety class (classification designation) of the system components (equipment, valves, pipelines, measuring circuits) in accordance with the RD requirements.

12.1.1.2. Input data

The following shall be provided:

- the list of safety RDs containing the requirements the system shall comply with;

- the input data for the design defining the required characteristics and parameters of the system and the external conditions where these characteristics shall be obtained;

- data on the ambient parameters in the room guaranteeing normal functioning of the components.

12.1.1.3. Design principles

It should be demonstrated that the system design is based on the principles and criteria including the safety assurance principles:

a) single failure principle. It should be demonstrated that the system is designed with due regard for the single failure principle;

b) redundancy principle. It should be demonstrated:

- that the design provides for redundancy of the individual system components in order to enhance reliability of the systems;

- how the expected downtime periods due to maintenance, testing and repair with the NPU in operation are taken into account in the analysis of reliability and redundancy adequacy for the systems performing safety functions;

c) diversity principle. It shall be demonstrated how the diversity principle is implemented in the design of systems and components in order to prevent common cause failures;

d) separation principle. The physical barriers separating the system channels or spatial separation aimed to prevent common cause failures (fires, flooding, etc.) shall be indicated.

The ways to activate (control) the system components as well as the signals used to activate the systems, the required energy and working medium sources shall be listed.

12.1.1.4. Requirements for associated systems

Requirements for the systems supporting the functioning of the safety system and interfaced with it by the steam, water, power supply, ventilation systems, etc. shall be specified.

12.1.1.5. Layout requirements

Requirements for location of the equipment and its mutual arrangement, routing of the pipelines and other components shall be specified.

12.1.2. System design

12.1.2.1. Description of the process flow diagram

The following information shall be provided:

a) configuration and basic technical characteristics of the system and its components;

b) description of the design and (or) process flow diagram of the system in general, its sub-systems and components (if they perform independent functions);

c) information on heat removal from the systems: characteristics of heat emissions, cooling media; supply of the media, characteristics of mechanical impurities;

d) information on water purification from radioactive substances and mechanical impurities:

purification means, water exchange rate, measures aimed to prevent clogging of the system components and loss of their heat transfer and throughput properties (fouling of heat exchangers, filters, screens, etc.).

It should be demonstrated that the system components withstand the ambient parameters (pressure, temperature, ship rolling, heel and trim difference, vibration, impact loads, humidity and radiation fields occurring in the course of operation) affecting the system components in all operation modes.

12.1.2.2. Substantiation of the required characteristics and parameters

The following characteristics and parameters of the working media shall be substantiated:

- flow rate, pressure, temperature, volume of the tanks, boric acid concentration, flow resistance of the pipelines;

- basic characteristics of the valves including dynamic characteristics (response time, operation principle);

- redundancy of the active system components, energy sources, instrumentation, etc. (in case of necessity these characteristics may be specified in other sections).

12.1.2.3. Description of components

The components shall be described in accordance with the standard description structure given in Subsection 12.2.

12.1.2.4. Description of the used materials

The applied materials of the system equipment and pipelines shall be described, including substantiation of the material selection with due regard for normal operation conditions, operational occurrences and design basis accidents; information on validation and experimental substantiation of the materials shall be provided.

12.1.2.5. Overpressure protection

The way to arrange overpressure protection for the system (in all operation modes) ensuring operability maintenance for the system components shall be demonstrated, and its characteristics shall be specified.

12.1.2.6. Layout of the equipment

The following information shall be provided:

- drawings and figures illustrating location of the components and operation of the system, its spatial layout and interfaces with other NPU systems in sufficient detail;

- location of the system equipment including radiation protection;

- access to the components for state monitoring, inspections, maintenance, testing, decontamination and repair; exposure doses for the personnel in the course of maintenance shall not exceed the established limits;

- consideration of the design load conditions (dead weight, dynamic impacts, thermal expansion in steady and transient modes);

- description of all limit stops, supports and shock absorbers installed on the pipelines and

equipment;

- accommodation of thermal movements;

- it should be demonstrated that location of the components complies with the requirements for the system layout.

12.1.2.7. Activation and disabling of the system

The list of postulated initiating events requiring activation of the system shall be provided. The system functioning algorithm shall be described.

12.1.2.8. Protection of the system against fires and flooding

Means and techniques for protection of the rooms where the system components are located against fires and flooding shall be specified.

12.1.2.9. Power supply of the system

The following information shall be provided:

- the power supply scheme (abridged) for the system components (consumers);

- distribution of the needs among the systems and power supply categories;

- schedule for connection of the consumers in the course of the system activation;

- permissible deviations with regard to the activation time, frequency and voltage.

12.1.2.10. Compressed air supply

The following information shall be provided: the compressed air supply scheme (abridged) for the system components; air flow rate, parameters and quality; description of the system operation, including its functioning in case of any failures.

12.1.2.11. Oil supply characteristics

The following information shall be specified for each oil consumer: oil flow rate, volume, parameters and quality; the oil replacement and removal procedure.

12.1.2.12. Water supply characteristics

Water reserve in the filling and make-up tanks shall be specified. The following information shall be provided for make-up: flow rate, parameters, permissible water supply interruptions.

12.1.2.13. Ventilation of the rooms

The following information shall be provided for the rooms (where the system components are located): characteristics of fans and ventilation systems, heat emission amounts, quantity of the supplied gases and aerosols, air exchange rate.

12.1.2.14. Gas removal and gas vents

Gas removal and gas vents from the system in the course of its functioning shall be specified.

12.1.2.15. Fire protection means

Fire protection means (fire detection and annunciation means, application of the relevant

materials for manufacturing of the system components, etc.) shall be described.

12.1.3. Control and monitoring of the PSS system operation

12.1.3.1. Measurement points

The list of the controlled system parameters (including instrumentation redundancy) shall be presented in the tabular format.

Parameter Measurement point index

Characteristics of the controlled parameter

Recommended sensor type

Information presentation (place, format)

Actions according to setpoint

Note

Rated value Setpoint value

12.1.3.2. Control and adjustment characteristics

The following information shall be provided:

- the list of valves, mechanisms of the PSS system and their operation algorithms in tabular format;

Mechanism

Mechanical valve drive type

Parameters of the pumped medium

Quantity

State of the valve or mechanism

Nature of control

Information presentation (place, format)

Note

- availability of the operator support means in the control of systems and components;

- habitability of the system control stations.

12.1.3.3. Information provided to the operator by the CCR display means

The following information shall be provided to the operator:

- availability (or unavailability) of the PSS for performance of the safety function;

- exceedance of the established operation limits for the operating parameters;

- any deviations from the normal PSS operation conditions in the course of the NPU functioning;

- the need to perform safety functions;

- performance of the safety functions by the PSS;

- the safety functions are performed or any failure in performance of the functions occurred.

12.1.4. Tests and inspections

12.1.4.1. Quality assurance for the PSS system

The basic requirements to quality assurance for the system and its components in the course of

design, manufacturing, construction, installation and operation in accordance with the QAPs for design, development, testing, construction and commissioning shall be specified.

12.1.4.2. The list of potential nuclear-hazardous works

The list of potential nuclear-hazardous works in the course of installation, testing, operation, repair and decommissioning of the system and its components shall be provided.

12.1.4.3. Testing scope and methodologies

The scope and methods of inter-departmental, commissioning tests, tests and inspections in the course of operation and their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation and hardware used for the purpose of testing shall be presented and substantiated.

12.1.4.4. State monitoring and testing of the system

Information on the methods, scope and time limits for state monitoring and testing of the system in the course of the RP operation, characteristics of the arrangements provided in the design for these purposes shall be presented, and their compliance with the RD requirements shall be demonstrated.

The hydraulic testing scheme for the system and its parameters shall be specified.

12.1.4.5. Maintenance and inspection procedure

The procedure for maintenance and periodic trial runs of the systems and (or) their individual components shall be provided.

12.1.4.6. Diagnostics of the system

Methods and techniques for monitoring of vibrations, noise, leaks and any other hidden defects shall be presented.

12.1.5. PSS design analysis

12.1.5.1. Reliability parameters of the system

Quantitative analysis of the system reliability with due regard for independent failures of its components, common cause failures and human errors shall be provided.

12.1.5.2. Normal operation

Functioning of the system and the associated actions of the operator (personnel) under normal operation conditions shall be described.

12.1.5.3. Failures in the course of normal operation

Failures of the system components including human errors and impacts of failure consequences (particularly common cause failures) on operability of the system, any interfaced systems and the NPU safety in general shall be analyzed. Failures requiring special consideration in Section 15 shall be specified.

12.1.5.4. Functioning of the system in case of accidents

Functioning of the system in case of design basis accidents with due regard for any failures of the system components including human errors and common cause failures shall be described.

The delay time within which any erroneous actions of the operator under emergency conditions will not result in any hazardous consequences shall be specified. The way to protect the systems against unauthorized intervention of the operators shall be demonstrated.

The possibility to use the system and its components to manage beyond design basis accidents shall be demonstrated.

12.1.5.5. Safety analysis of the PSS design

The list of calculation programs used to analyze safety of the system, information on validation and verification of the calculation programs shall be provided. The scope of information shall be sufficient to perform independent alternative calculations in case of necessity. Information on the calculations shall be grouped into the following groups:

- thermal and hydraulic calculations;

- strength calculations for the system components;

- calculations of the radiation situation after the PSS activation;

- calculations of reliability parameters for the system and its components.

Information on the calculations shall consist of the following parts:

- the list of all calculations performed;

- the list of methodologies and programs used for safety analysis with indication of the application scope, assumptions, information on validation of the programs;

- analysis of the calculation results;

- conclusions.

In case any experiments were performed in order to substantiate safety of the system design the conditions of the experiments shall be described, their compliance with the design conditions shall be analyzed, the experimental facilities and metrological support of the experiments shall be described, and interpretation of the results with regard to the design conditions shall be provided.

Additional information:

- the list of all experimental works performed;

- analysis of the experiment results with conclusions;

- the list of works to be performed at the detailed design development stage.

12.1.5.6. Comparison with similar PSS designs

Results of comparative analysis with similar PSS designs shall be provided, characteristic drawbacks of the previously used systems and any engineering arrangements applied in this PSS shall be specified.

Each subsection of Section 12 shall be completed with the analysis of compliance with the requirements, principles and criteria of the relevant regulatory documents on safety. In the course of information presentation references to other sections where this information is provided in more details are possible.

Particular content of each subsection may change depending on the system peculiarities. Individual subsections may be omitted or supplemented with other ones if it is determined by the system peculiarities.

12.1.5.7. Conclusions

Conclusions on the system compliance with its functional purpose and the RD requirements for safety shall be made.

12.2. Description structure for the components of protectivesafety systems

12.2.1. Design basis

12.2.1.1. Purpose and functions

The following shall be specified:

- purpose and functions of the component;

- additional safety and normal operation functions performed by the system components;

- safety class (classification designation) of the system component in accordance with the RD requirements.

12.2.1.2. Input data

The following shall be specified:

- the list of regulatory documents containing the requirements the described component shall comply with;

- the input data for the design defining the required characteristics and parameters of the component and the external conditions where these characteristics shall be obtained;

- data on the ambient parameters in the room guaranteeing normal functioning of the component.

12.2.1.3. Design principles

The fundamental solutions used as the basis for the component design shall be specified.

12.2.1.4. Requirements for associated systems

Requirements for the systems supporting the functioning of the component and interfaced with it by the steam, water, power supply, ventilation systems and other media shall be specified.

12.2.1.5. Layout requirements

Requirements for the component installation shall be specified.

12.2.2. Component design

12.2.2.1. Description of the component structure

The following information shall be provided:

a) basic technical characteristics of the component;

b) description of the component structure;

c) drawings, figures and schemes illustrating the component design and operation in sufficient details;

d) the ambient parameters affecting the component in all operation modes. It should be demonstrated that the component is designed with due regard for its capability to withstand the ambient conditions (pressure, temperature, ship rolling, heel and trim difference, vibration, impact loads, humidity and radiation fields occurring in the course of operation). It should be noted that the component shall withstand these conditions during design basis accidents and after them and also within the entire service life of the component;

e) information on heat removal from the component: characteristics of heat emissions, cooling media; supply of the media, characteristics of mechanical impurities;

f) information on water purification from radioactive substances and mechanical impurities: purification means, water exchange rate, measures aimed to prevent clogging of the component and loss of its heat transfer and throughput properties (fouling of heat exchangers, filters, screens, etc.);

g) characteristics of oil supply to the component (in case of necessity): oil flow rate, volume, parameters and quality; the oil replacement and removal procedure;

h) maintainability;

i) labeling, painting, corrosion protection;

j) the manufacturing plant.

12.2.2.2. Description of the used materials

The applied materials of the component shall be described, including substantiation of the material selection with due regard for normal operation conditions and operational occurrences including emergency situations and accidents; information on validation and experimental substantiation of the materials shall be provided.

The material shall be presented in the following sequence:

- information on structural and welding materials as well as on their compatibility with the process media;

- substantiation of selection of the materials with due regard for normal operation conditions and operational occurrences including emergency situations and accidents, information on validation of the materials and their experimental substantiation;

- information on metal state monitoring for the component;

- information on the methods, scope and time limits for state monitoring and testing of the component in the course of the RP operation, characteristics of the arrangements provided in the design for these purposes and their compliance with the RD requirements.

12.2.2.3. Overpressure protection

12.2.2.4. Location of the component

Additional information on the system components with due regard for peculiarities of these

components shall be presented.

1. Pipelines and their components:

- the list of pipelines

- design, location, layout, routing conditions, inclinations;

- design of supports, fasteners, suspenders, penetrations and expansion joints;

- drainage and air removal pipelines;

- welding information;

- permissible warm-up and cooldown rates;

- information on safety devices;

- the main incoming control data in the course of manufacturing and installation (metal quality, welding quality, hydraulic testing results);

- the list of controlled parameters and scope of diagnostics in the course of operation (state of the basic metal and weld joints, movements and vibrations, erosion and corrosion wear, chemical composition of the media, heat insulation state);

- heat insulation design and calculations.

2. Valves:

- the list of valves;

- information on compatibility of the structural and welding materials with the process media;

- characteristics (leak-tightness, flow resistance, opening pressure for check valves);

- information on the drive;

- the drive parameters;

- response time;

- permissible pressure differential;

- layout, location, ambient conditions;

- design of supports and fasteners;

- permissible warm-up and cooldown rates;

- the list of parameters controlled in the course of operation and scope of diagnostics (movements, vibration, wear, leak-tightness, parameters of the drive).

3. Heat exchangers:

- the list of heat exchangers;

- thermal calculations;

- characteristics: the working and cooling media, parameters of the media (pressure, temperature,

flow rates and velocities), heat transfer coefficient, flow resistance of the circuits, protection and interlocking devices, layout conditions;

- requirements for the cooling water quality;

- information on instrumentation and controls;

- design of supports and fasteners, permissible warm-up and cooldown rates;

- the list of parameters controlled in the course of operation and scope of diagnostics (movements, vibration, leakages, media parameters, characteristics of mechanical impurities in the media, changes of heat transfer coefficients);

- heat insulation design;

- overpressure protection (scheme, design and characteristics of protective devices, calculation and experimental substantiation of their operability);

- the technique for tube leakage detection and elimination of defects;

- the technique for the heat exchange surface cleaning.

4. Pump sets:

- the list of pumps;

- characteristics: capacity, pressure head, power, time of turn, net positive suction head, starting current of the motor, suction lift, information on suction vortex formation, requirements for the water purity from any mechanical impurities, vibration characteristics, temperature of the pumped water;

- permissible number of start-ups per an hour;

- information on instrumentation and controls;

- protections and interlocks;

- layout and location conditions;

- design of supports and fasteners;

- ambient conditions (temperature, humidity);

- the list of parameters controlled in the course of operation and scope of diagnostics (movement, vibration, leaks from sealing glands, water and oil parameters, pump characteristics).

5. Tanks:

- the list of tanks;

- characteristics: - volume, medium exchange rate;

- drainage and air removal;

- assurance of uniform poison concentration within the tank volume;

- sludge removal technique;

- assurance of the design process medium level and overflow prevention;

- layout conditions;

- design of supports and fasteners;

- the list of parameters controlled in the course of operation (levels, permissible leakage value, media parameters, poison concentration).

12.2.3. Control and monitoring

12.2.3.1. Measurement points

12.2.3.2. Control and adjustment characteristics

12.2.3.3. Operator's actions

12.2.4. Tests and inspections

12.2.4.1. Quality assurance requirements

The basic quality assurance requirements for the component in the course of manufacturing, construction and installation shall be provided in this subsection.

12.2.4.2. Control scope and methods

The scope and methods of the incoming control, inter-departmental, commissioning tests, tests and inspections in the course of operation and their metrological support shall be substantiated; the list and permissible values of the controlled parameters and the requirements for the instrumentation and hardware used for the purpose of testing shall be presented and substantiated.

The following information on the component at the operation stage shall be provided:

- methods, scope, time limits for state monitoring and testing of the component;

- schedules of maintenance and periodic inspections of the component;

- information on the stock of consumables, spare parts and assemblies.

12.2.4.3. State monitoring and testing

Information on the methods, scope and time limits for state monitoring and testing of the component in the course of the RP operation, characteristics of the arrangements provided in the design for these purposes and their compliance with the RD requirements shall be presented.

Methods and means for monitoring of vibration, noise and leakage shall be provided within the framework of the component diagnostics.

The hydraulic testing scheme for the system and its parameters shall be specified.

12.2.4.4. Maintenance and inspection procedure

The procedure for maintenance and periodic trial runs of the components shall be provided.

12.2.5. Design analysis

12.2.5.1. Reliability parameters of the component

Reliability parameters of the component shall be presented (the design service life, the assigned operation time, period of continuous operation, etc.).

The analysis results shall be provided to confirm that functional reliability of the components performing safety functions complies with the requirements for characteristics in accordance with the assumptions used for the analysis of initiating events. Information on the possibility of inspection (diagnostics) shall be provided for the components referred to the second safety class.

12.2.5.2. Normal operation

Functioning of the component under normal operation conditions shall be described.

12.2.5.3. Functioning of the component in case of failures

Any failures of the component parts and impact of the failure consequences (including common cause failures) on the operability of the component under consideration shall be analyzed.

12.2.5.4. Safety analysis

12.2.5.5. Comparison with similar designs

12.2.5.6. Substantiation of any deviations from the requirements of regulatory documentation

12.2.6. Conclusions

Conclusions on compliance of the component with the RD requirements for safety shall be made.

12.3. Localizing safety systems

12.3.1. General description

12.3.1.1. Purpose and design basis

All LSSs and components performing the confinement functions shall be listed.

The following shall be provided:

- purpose of each system, safety classes shall be specified;

- principles and criteria used as the basis for the system design including the requirements from the RP;

- load limits for the LSS components due to design basis accidents and external impacts typical for the ship, permissible values of the reliability parameters;

- information specifying how and to what extent the possibility for state monitoring, maintenance, testing, repair and decontamination of LSSs and their components is provided;

- experimental substantiation of operability for LSSs and their components, description of the testing facilities, the methodology of experiments and the basic results of the experiments, experimental substantiation of all LSS operation modes;

- calculations confirming capability of the LSS components to accommodate loads from the postulated design basis accidents and external impacts within the design limits and in combinations defined in the regulatory documents without any destruction and loss of operability;

- the input data for these calculations, the basic assumptions for development of the calculation algorithms and algorithms themselves within the scope enabling any independent expert to repeat these calculations, information on testing, verification and validation of the applied calculation

programs;

- confirmation of the fact that all LSSs and their components will withstand the number of in-house tests provided in the design and also the required number of loading cycles with excessive pressure and underpressure during strength and tightness testing of the containment and protective enclosure in the course of commissioning works and operation without any loss of operability;

- the time period from the beginning of a design basis accident with loss of coolant and up to the moment when the personnel access to the accident localization area becomes possible. The same time period shall be substantiated for beyond design basis accidents;

- information on the way to control and monitor the active LSS components;

- analysis of the need, the scope of control and monitoring of the active LSS components from the CCR; passive components with mechanical moving parts from the CCR, the ECS or any other rooms and local stations; in this case it should be demonstrated that the requirements for performance of the functions for limitation of RSb releases into the environment by these components are taken into account;

- measures aimed to prevent harmful impact of microorganisms on the LSS components contacting with solutions in the course of normal operation.

12.3.1.2. Description of the design and (or) the process flow diagram

The following shall be specified:

- description of the design and (or) the process flow diagram of the system with indication of the systems and components performing independent functions, including fasteners, supports, foundations, etc. Descriptions of individual components may be provided in separate subsections with the same structure as for the description of the system in general;

- detailed figures and schemes illustrating the design or the process flow diagram of the system and the basic technical characteristics of the system and its components.

12.3.1.3. Control and monitoring of the system

Control and monitoring shall be described for each system, and characteristics of the parameters (setpoints) for activation of the process protections and interlocks shall be specified.

12.3.1.4. Materials

Selection of the materials shall be substantiated with due regard for the normal operation conditions and any operational occurrences including accidents.

12.3.1.5. Quality assurance in the course of manufacturing, installation and construction

Quality assurance programs for all LSS components in the course of manufacturing, installation and construction shall be provided.

12.3.1.6. Control and testing in the course of operation

Information on the methods, scope and time limits for state monitoring and testing of the system in the course of the ship operation, characteristics of the arrangements provided in the design for these purposes shall be presented, and their compliance with the RD requirements shall be demonstrated.

12.3.1.7. Functioning of the system

Functioning of the system shall be described, particularly with due regard for any potential failures in other systems of the RP (in accordance with the single failure principle), and characteristics of the measures provided in the design in order to protect the system against any impact of these failures shall be presented.

The main characteristics (mechanical, thermohydraulic, physical and chemical, strength) and reliability parameters shall be specified for each system operation mode including failures of other systems; it should be also demonstrated that they do not go beyond the permissible limits determined in par.12.2.1.1.

12.3.1.8. Functioning of the system in case of failures

Failures of the system components including human errors (in accordance with the single failure principle) shall be analyzed, and the impact of failure consequences on the system operability and safety of the ship in general shall be assessed.

In this case failures of passive components with mechanical moving parts (for example, check valves), active components (gate valves, pumps, etc.), instrumentation of the system itself and the associated CSS and SSS shall be considered. Special attention shall be paid to analysis of common cause failures including any potential fires.

Qualitative and quantitative characteristics of consequences shall be provided for the failures under consideration, including changes of the main parameters affecting the ship safety; impact of these failures on operability of other systems shall be demonstrated.

12.3.1.9. System reliability analysis

Qualitative and quantitative reliability analysis for the system shall be provided based on the data specified in par. 12.2.5.1.

It should be demonstrated that the unavailability factor for the LSSs in the course of the NPU operation does not exceed the established value.

12.3.1.10. System design assessment

It should be demonstrated on the basis of the analysis that the system design complies with the adopted safety requirements, principles and criteria.

12.3.2. Containment and protective enclosure

The main components of the containment and the protective enclosure shall be listed in addition to par. 12.2.1.1 - 12.2.1.5.

It should be demonstrated that:

- structures of the containment and the protective enclosure perform their functions in accordance with the RD requirements;

- the containment is designed for the internal pressure caused by emergency release of the coolant in case of the primary circuit rupture;

- leak-tightness of the containment complies with the RD requirements;

- information on the means for testing of the containment under excessive pressure in the course of regular examinations and after the nuclear core refuleling is presented in the design.

12.3.2.1. Hatches, airlocks, doors

The following shall be specified:

- basis and purpose for selection of the certain containment components, conditions taken into account in order to define the number of hatches, airlocks or doors;

- purpose of each hatch, airlock or door and the established requirements for their leak-tightness as well as the relevant drawings;

- method and frequency of leak-tightness checks for the rooms equipped with hatches, airlocks and doors in the course of operation as well as their accessibility;

- that the design of hatches, airlocks and doors ensures the leak-tightness degree and ionizing radiation attenuation factor as per the design under normal operation conditions and in case of any design basis and considered beyond design basis accidents;

- permissible leakage through the hatches, airlocks and doors under the design pressure;

- opening direction of the doors, availability of position indication for the hatch covers and door leaves (sealed, unsealed) of the CCR and the ECS and mechanical or electrical interlocks preventing simultaneous opening of both airlock doors;

- whether the airlock doors are equipped with pressure equalizing valves with indication of their position;

- the possibility to activate opening-closing mechanisms of the door leaves and hatch covers manually by one person both from outside and from inside the containment or the lock;

- how the design of airlocks enables urgent evacuation of the personnel in emergency situations;

- information on emergency lighting and two-way communication system between the locks and the CCR and ECS;

- the standards used to perform strength calculations for the hatches, airlocks and doors and their embedded parts;

- elevations where any airlocks, hatches and doors used for evacuation are installed in relation to the floors of the rooms. Potential water level on the floor occurring in case of any accidents shall be also specified here.

12.3.2.2. Penetrations

The following shall be specified:

- all types of penetrations, their diagrams and (or) drawings;

- methods to control weld joints in the course of manufacturing, after installation and during operation;

- permissible value of leakage through each penetration under the design medium pressure in the containment;

- description of the way to arrange group electrical penetrations taking into account the physical separation principle for safety channels.

12.3.2.3. Isolating devices

All pipeline networks crossing the containment and indicated on the relevant schemes shall be

listed. Any medium the pipelines are connected to inside and outside the containment as well as the number of isolating devices and their installation places shall be indicated on the same schemes; principles for installation of the isolating devices on the networks crossing the containment shall be defined and specified.

The list of networks where installation of isolating devices is not mandatory shall be provided; in this case the relevant substantiations shall be presented.

Calculations used as the basis to select the type of isolating devices and to take into account their response time, the input data for the calculations, claculation methodologies and programs shall be provided.

The following shall be specified:

- the list of initiating events when the main pipelines crossing the containment shall be isolated. Time dependence of the discharge in case of any isolating device failure shall be specified for each main pipeline;

- the regulatory documents containing the requirements that the pipeline valves used as isolating devices shall comply with;

- permissible leakage value under the design pressure for all types of isolating devices and the total number of devices referred to each type;

- frequency of testing for the isolating devices with pneumatic and (or) electric drives (if any);

- types of valves that cannot be used as isolating devices;

- means and measures provided in the control systems of the isolating devices in order to prevent any unauthorized opening or closure resulting in release of radioactive substances or damage of essential ship components and systems both at the moment of accident and during the post-accident period;

- that the NPU safety level will not be decreased during trial run of the isolating devices on the course of the reactor power operation either individually or within the SS channel (in case such trial run is provided in the design).

12.3.2.4. Bypass and safety devices

The following shall be demonstrated:

- where and for what purpose the bypass and safety devices are applied (including the devices for equalizing of the external and internal pressure in case of the ship drowning) and how they function;

- when the containments without any regular safety devices shall be equipped with such devices (for example, within the period of the containment strength and leak-tightness testing);

- that safety devices ensure leak-tightness of the room under the design basis accident conditions;

- how the quantity and throughput capacity of the safety devices are selected (the conditions used as the basis for selection);

- that the design of safety and bypass devices enables to perform individual testing for response and leak-tightness as well as replacement of sealing components, inspection and repair with the reactor shut down;

- that techniques and methods are provided to perform periodic testing of the safety and bypass

devices for response and operability.

12.3.3. Systems of pressure reduction, heat removal, hydrogen removal and gas-aerosol treatment

12.3.3.1. Passive steam condensers

The following shall be demonstrated:

- the main components of passive steam condensers and the relevant drawings;

- that passive condensers of steam generated in the course of accidents with the primary circuit integrity loss have sufficient reserve of cooling medium in order to ensure reliable condensing of all generated steam. Otherwise it should be demonstrated that the tanks of passive condensers are equipped with pumping and heat exchanging units with sufficient capacity and the required redundancy;

- the requirements to be followed in the design of the passive steam condenser partitions in case they form part of the containment or the protective enclosure and also in case the condensing devices are installed in tanks;

- that inlets and outlets of the steam supply channels are free from any pipelines and equipment. Otherwise its should be demonstrated that these components and their fasteners are designed for the impact of steam-air mixture stream and any other potential dynamic impacts, and the area of the section free from equipment and pipelines is sufficient to ensure non-exceedance of the design parameters inside the containment in case of any accidents with loss of coolant;

- filling and emptying systems for the tanks, water purification systems for the tanks, level and temperature control systems;

- that passive steam condensers retain their operability under design values of rolling, heeling and trim difference of the ship;

- the medium parameters (pressure, pressure differential, temperature and humidity) the passive steam condenser tanks are designed for with due regard for its dynamic impact;

- the way to prevent any damage of the walls and covers of the passive steam condenser tanks due to hydraulic shocks possible because of steam-gas mixture agitation as well as potential vacuuming of the containment in case of accidents or spurious actuation of the spray system;

- accessibility of the tank surfaces for repair and inspections;

- experimental substantiation of operability for the passive steam condensers; in this case all possible modes of their operation shall be characterized.

12.3.3.2. Passive spray devices

The following shall be specified:

- the main components of passive spray devices and the relevant drawings;

- accessibility of the PSD tank surfaces for inspections and repair;

- the filling and drainage systems for the PSD tanks and devices for monitoring and measuring of the water level and temperature in the tanks;

- requirements for leak-tightness of the PSD siphon pipes and their tightness control, experimental substantiation of the PSD design operability as well as all possible operation modes;

- the requirements used as the basis to define the composition of the solution sprayed by the PSD, measures aimed to prevent non-homogeneity of the solution within the tank volume and means for purification and chemical composition adjustment.

12.3.3.3. Active spray system

The following shall be specified:

- the main components of active spray devices and the relevant drawings;

- the requirements taken into consideration to define the chemical composition of the solution sprayed by the system, measures aimed to prevent non-homogeneity of the solution within the volume of the spray system tanks and means for the solution purification and chemical composition adjustment.

- that the active spray system is designed and manufactured in such a way so that it could be tested under the conditions as close as possible to the emergency ones and all operations activating the system including switching to the emergency power supply source could be obtained in practice;

- experimental operability substantiation for all components of the spray system in all possible operation modes;

- that any hazardous impacts on the equipment related to operation of the spray system in the course of testing are minimized, and the possibility to check operability of the active spray system components, including the spray pumps, in provided in the course of the reactor power operation;

- the way to arrange control of the active spray system from the CCR and the ECS in case of different accidents;

- availability of shut-off devices on the spray system pipelines regardless of the drive type, position indication in the CCR and the ECS;

- the way to prevent loss of the containment leak-tightness via the spray system pipelines in case of any failure to start the spray pump upon an emergency signal;

- systems for monitoring of thermal and technical parameters of the active spray system (pressure, temperature, flow rate) with indication of the instruments and sensors as well as for monitoring of the chemical parameters of water sprayed inside the containment.

12.3.3.4. Containment ventilation system

It should be demonstrated that:

- the air ducts of the containment ventilation systems are isolated quickly and reliably through the use of shut-off valves under normal operation conditions and in case of any operational occurrences and design basis accidents;

- ingress of condensate or moisture into the equipment located inside the containment is prevented in case of the system use for functioning under normal NPU operation conditions;

- systems for monitoring of the parameters and control of operation for the ventilation system components performing the LSS functions are linked with the CCR and the ECS;

- operability of the ventilation system design is substantiated by experiments for all operation modes.

12.3.3.5. Hydrogen concentration monitoring and emergency removal system

The following shall be demonstrated:

- the points of the containment rooms where hydrogen concentration monitoring is provided and the place where the information on concentration is transmitted. Location of the hydrogen concentration monitoring points shall be substantiated;

- how and from where the emergency hydrogen removal system is controlled;

- the alarm means activated in case of any exceedance of the hydrogen concentration in the containment established in the design.

The following shall be specified:

- information on the materials located inside the containment (heat insulation, chemical coatings, etc.) capable to be engaged in any hydrogen-generating chemical reactions with the media in case of any accidents with coolant loss;

- calculations to substantiate hydrogen accumulation with due regard for all processes inside the containment, performance of the assigned functions by the emergency hydrogen removal system in case of design basis accidents;

- experimental substantiation of operability for the emergency hydrogen removal system with due regard for all possible operation modes.

12.3.3.6. Emergency gas and aerosol treatment plants

It should be demonstrated that:

- the filtering elements of the emergency gas treatment plant are accessible under normal operation conditions and in the post-accident period for their replacement, and the required degree of leak-tightness and biological protection is ensured for these components;

- operation of the plants is efficient, and the experimental substantiation results for their design take into account all possible operation modes.

12.3.3.7. System of passive heat removal from the containment

Drawings of the PHRS design and the relevant explanations, results of the experiments to substantiate operability of the PHRS design or the relevant calculation substantiation for all possible operation modes and the CW results shall be provided.

12.3.3.8. Any other LSSs may be present at the RP; they shall be also described in ths SAR in accordance with the SAR section "General provisions".

12.3.4. Testing of the LSSs and their components

It shall be demonstrated how the LSSs and their components shall undergo verification for compliance with the design characteristics after manufacturing, in the course of commissioning, after repair and regularly within the entire service life of the ship.

The following shall be specified:

- types of tests for the LSSs and their components with regard for compliance with the design characteristics as well as validated testing methods;

- the list of documents used as the basis for testing of the LSS components after their manufacturing, installation and in the course of operation and the persons performing the testing;

- time limits and methods for testing of the containment, the protective enclosure and their components for strength and leak-tightness after installation and operation.

Devices and (or) systems required to perform testing shall be listed.

The following shall be specified:

- when, how and for what purpose the functional testing of the LSSs and their components is performed;

- particular parameters checked in the course of functional testing, the procedure for the personnel admittance to performance of the tests;

- the procedure for the personnel admittance to inspection of the structures in the course of pressure or load increasing;

- the places where the personnel and control instruments engaged in the testing shall be in the course of load increasing and reduction;

- any actions of the personnel prohibited in the course of testing, as well as actions of the personnel after detection of any defects.

12.3.4.1. Strength testing of the containment in the course of the ship construction

The following shall be specified:

- the testing procedure;

- strength assessment criteria;

- information on design of the sensors for measurement of the stress-strain parameters and their errors.

12.3.4.2. Leak-tightness testing of the containment

The way (signals) to close the isolating devices on the networks crossing the containment in the course of leak-tightness testing shall be demonstrated.

The method used to define the degree of the containment leak-tightness shall be described. It should be demonstrated that this method corresponds to the leakage value determination accuracy, requires minimum time for testing with this leakage value and is validated in accordance with the established procedure.

The testing methodology shall be described, and safety measures implemented in the course of testing shall be specified.

The following shall be specified in the testing methodology:

- when and how closed position of the manual isolating valves will be ensured;

- when and how the air-driven and motor-driven isolating valves will be closed;

- what technical devices will be used to build excessive air pressure inside the containment;

- the criterion used to determine stabilization of the parameters inside the containment;

- recording frequency for the parameters;

- duration of the containment holding under pressure;

- where and how the detected containment defects are recorded;

- the number of testing pressure stages for the containment leak-tightness testing in the course of commissioning works;

- the criterion for assessment of the containment leak-tightness testing results in the course of pre-operational commissioning works under the design and reduced pressure and in the course of operation under the reduced pressure;

- pressure increase and decrease rate inside the containment in the course of leak-tightness testing.

The leakage value calculation algorithm in the course of the containment leak-tightness testing shall be provided.

12.3.4.3. Leak-tightness testing of the containment components

All containment components subject to leak-tightness testing shall be listed.

Drawings enabling to understand the design of each containment component subject to the testing as well as the testing methodology, criteria of successful testing completion both in the course of the ship construction and commissioning works and in the course of operation shall be provided.

The following shall be specified:

- time of testing;

- requirements for the containment components with regard to their accessibility for testing;

- the testing program for the containment components in the course of commissioning works;

- the scope of incoming control and post-installation testing and also acceptance criteria for the components;

- frequency of testing for the containment components in the course of operation and criteria for performance of unscheduled testing.

12.3.4.4. Hydraulic testing of rooms and tanks

The following shall be specified:

- the rooms and tanks representing the LSS components and subject to hydraulic testing;

- time of testing;

- hydraulic testing methodology;

- criteria for early termination of the tests as well as their successful completion.

12.3.4.5. Functional testing of the active spray system and water tanks of the spray system pumps

The following shall be specified:

- time limits for functional testing of the active spray system and water tanks of its pumps;

- the scope of checks in the course of testing and the testing methodology;

- successful testing criteria;

- frequency of testing;

- documentation used as the basis for testing of the active spray system components and water tanks of its pumps.

12.3.5. Servicing and maintenance of the LSSs in the course of operation

The following shall be specified:

- information on the documents containing the requirements to be fulfilled in the course of LSS maintenance, safety assurance in the course of LSS maintenance and keeping them in good operable condition;

- basic requirements for the LSS operation manual, information on the scope and frequency of LSS maintenance and operability checks, success criteria for the inspections.

The following shall be specified:

- frequency and types of operability checks for active and passive LSS components;

- the procedure for documenting of the inspection results;

- persons responsible for development of LSS operation manuals and engaged in their preparation, coordination and approval;

- state of the LSSs at any reactor power level;

- states of the LSSs when activation of the reactor is prohibited;

- LSS inspections subsequent to any repair works;

- LSS inspections prior to the reactor activation and the required documentation;

- the list of the LSS components with forbidden personnel access during power operation of the reactor;

- the list of the LSS components with restricted personnel access during power operation of the reactor;

- controlled parameters of the LSS systems and conponents in the course of the NPU power operation;

- time (with substantiation) required to recover the LSS operability upon expiry of which the reactor shall be brought to sub-critical state if their operability is not recovered;

- documentation issued upon completion of any repair works and functional checks of the repaired LSS component (and the entire LSS in case of necessity);

- information to be recorded in the LSS data sheet.

12.4. Supporting safety systems

Information on the following SSSs shall be provided in this subsection:

- compressed air and hydraulic systems (if any) used as the energy source for safety systems;

- fire extinguishing;

- supporting ventilation systems.

In case there are any other SSSs on the ship they shall be described in accordance with the requirements of the SAR section "General provisions".

All SSSs, their components provided in the design as well as references to any other SAR sections containing information on these systems shall be presented.

Description of each system shall be provided in accordance with the following structure:

12.4.1. Design basis

The requirements of par. 12.4.1 are similar to the requirements stated in par. 12.1.1.

Safety assurance principles and criteria provided in the system design shall be specified, and their fulfillment shall be demonstrated.

12.4.2. System design

The requirements of par. 12.4.2 are similar to the requirements stated in par. 12.1.2.

12.4.3. Control and monitoring of the system operation.

Control and monitoring shall be described for each system, and characteristics of the parameters (setpoints) for activation of the process protections and interlocks shall be specified.

12.4.4. Tests and inspections

The following information shall be provided:

- commissioning of the system including its testing; the objectives of the main CW stages shall be specified, and description of these stages with indication of the testing methods and parameters shall be provided;

- any works in the course of which safety incidents can occur and any arrangements preventing occurrence of accidents;

- safety substantiation for the pre-operational testing of the systems;

- control and testing in the course of SSS operation;

- methods, scope and time limits for state monitoring and testing of the SSSs in the course of the NPU operation, the arrangements provided in the design for these purposes and their compliance with the RD requirements.

12.4.5. Design analysis

The following shall be specified:

- qualitative and quantitative reliability analysis for the system in accordance with the RD requirements. It should be demonstrated on the basis of the analysis that the system design complies with the established safety requirements, principles and criteria;

- analysis of the system functioning under normal operation conditions, description of the system functioning similar to par. 12.1.5.

- analysis of the system functioning in case of any failures. Any failures of the system components shall be analyzed similar to par. 12.1.5 (safety analysis).

QAPs for all system components in the course of manufacturing and installation shall be described.

12.4.6. Additional information

The following information on the SSSs shall be provided:

a) concept used as the basis for the system design, particularly:

- the possibility to perform the function in any emergency situation including blackout;

- the possibility for control and testing in any normal operation modes without any loss of functional properties;

- duration (limited or unlimited) of functioning in the emergency period;

- combination of the functions of safety systems and normal operation systems without any safety deterioration;

- approbation of the design solutions;

- the design limits non-exceedance of which is ensured by the system;

- comparison with similar solutions available internationally;

- any permitted deviations from the requirements of safety standards and rules;

b) information on protection of the SSSs against fires, flooding, physical damage, mechanical impacts in case any accidents with pipeline ruptures;

c) information on the possibility for the system operation under beyond design basis accident conditions;

d) the procedure for the system maintenance and inspections;

e) information on the stock of consumables, spare parts, lubricants, cooling agents, etc.;

f) information on the system control:

- activation and disabling interlocks;

- activation delays;

- activation and disabling prohibitions;

g) system control functions performed manually:

- with time-limited operator intervention prohibition;

- without any time limitations;

h) permissible time for power supply of the system:

- the procedure for activation of the system and its components in the blackout mode in accordance with the step-by-step start-up program.

The following information shall be provided:

- the means for the operator's support in control of the system;

- characteristics of the local stations where the system and its individual components can be activated;

- state monitoring for the system equipment as well as control methods and means (metal control for the pipelines, components, state monitoring for the assemblies, electrical resistance);

- diagnostics of the systems, methods and means for control of vibration, noise and leak-tightness loss;

- heat removal from the system (heat emitted in the course of the equipment operation, heat removed by the system);

- hydraulic testing of the system (schemes and parameters of hydraulic testing);

- the system filling and make-up (volumes, flow rates in the course of filling and make-up);

- bracing of the system components (limit stops, supports, thermal expansion joints);

- stability of the applied materials and their coatings under normal operation and accident conditions. Special attention shall be paid to generation of any secondary decomposition products posing hazard from the viewpoint of toxicity and explosiveness under any system conditions different from the design ones. For example freon decomposition process in the cooling units in case of a fire shall be considered;

- consideration of the requirements for the ship NPU decommissioning;

- interfaces with other systems and requirements for other systems.

13. COMMISSIONING

This section of the SAR shall contain the information on arrangement, scope, sequence and time limits for the commissioning works and tests carried out in the course of the ship NPU commissioning and shall cover any equipment, systems and components of the ship NPU related to assurance of its safe operation.

The information shall cover all commissioning stages beginning from acceptance, installation of the systems and finishing with integrated testing of the NPU and the ship with the rated power in the course of sea acceptance testing and putting into operation (including such types of works as pre- and (or) post-installation cleaning of the equipment and circuits, functional adjustment and testing of individual equipment units and valves as well as the entire systems, integrated testing of the NPU equipment, the nuclear core loading, physical start-up and neutronic tests, thermotechnical and harbour acceptance tests in accordance with the programs).

The following shall be specified and substantiated in this section:

- the basic fundamental provisions of the ship NPU commissioning program with the success criteria for all its stages and sub-stages enabling to assess the possibility for successful completion of the entire set of commissioning works;

- the main administrative and technical QAP arrangements in the course of the NPU commissioning.

The information shall demonstrate that:

- the RD requirements are fulfilled to the full extent in the course of commissioning;

- safety is ensured in the course of commissioning works and testing at all stages of the NPU commissioning;

- the required completeness of investigations and inspections is ensured for all modes and characteristics of the systems and the entire NPU related to its safe operation assurance.

13.1. Requirements for the information introduced to the SAR at the NPU commissioning stage

13.1.1. General provisions

The basic provisions of the NPU commissioning programs and quality assurance programs in the course of commissioning shall be defined and substantiated, including division of the works into stages and sub-stages, their interrelation, the procedure and time limits for completion of each stage or sub-stage, criteria of their successful completion, the required administrative and technical arrangements.

Compliance with the following basic conditions shall be demonstrated in presentation of the information on the NPU commissioning:

a) the organizational structure is arranged in order to ensure management of commissioning works and testing, in-process analysis of their results and timely adjustment of the work program (in case of necessity);

b) the sequence of works for inspections, adjustments and tests during commissioning of systems and components is optimized with regard to safety conditions, fabricability, resource management, independence and safety at any moment of testing from non-tested systems and (or) their equipment;

c) inspections are provided and the design characteristics of the systems and components including safety systems and safety-related systems are confirmed (with documenting);

d) the following is provided to the full extent:

- adequacy checks for the operational and emergency guidelines, process restrictions, safe operation limits and conditions;

- implementation checks with regard to the arrangements for maintenance of the systems and components;

- timely arrangement of accounting for operation modes and loading cycles for the equipment with the useful life limited with regard to cyclic strength and durability;

- performance of the warranty obligations by the suppliers and manufacturing plants;

e) the personnel have the required skills for the NPU control, operation and maintenance. The knowledge check shall be performed before admittance to work.

13.1.2. Organization of works

The expected arrangement of the works and structure of interaction between the operating organization (license holder) personnel and representatives of any scientific, design, engineering, installation, construction and commissioning organizations as well as the supplier organizations and inspectors of the state safety regulating authority both in the course of preparation for commissioning

and the NPU commissioning shall be described.

Distribution of the managing and executive functions and responsibilities between the organizations involved in the works and between the executives at various levels aimed to achieve the commissioning goals shall be demonstrated. Arrangement of the works, engagement of any companies in commissioning and selection of the personnel shall comply with the RD requirements.

The following shall be reflected in the presented information:

- the organizational structure of the operating organization, the NPU and ship personnel, their rights and liabilities, requirements for qualification;

- the administrative measures implemented by the operating organization, the NPU and ship designers, suppliers of the equipment and any other organizations engaged in performance of the works (arrangement and organizational structure of the state acceptance commission);

- functions of the organizations engaged in the ship NPU construction, their interaction and subordination, distribution of liabilities and responsibilities, as well as the requirements for the personnel qualification (brief characteristics of the composition, functions and working principles of the organizations shall be provided with the reference to the relevant documents);

- plans for involvement of additional personnel for each commissioning stage, the requirements for their qualification;

- administrative measures for safety assurance including development of the list of potentially hazardous works and technical requirements for their performance.

13.1.3. Stages of works

Division of the NPU commissioning process into stages and sub-stages shall be substantiated with due regard for the peculiarities of each particular NPU and the tasks solved at each stage (sub-stage); content of the main commissioning stages shall be specified, selection of the optimal sequence of works, performance and (or) combination of tests shall be substantiated, the arrangements aimed to control their performance shall be specificated, and the acceptance criteria shall be defined.

Information on the following stages shall be provided:

- adjustment and commissioning of the set of systems supporting acceptance, preparation and filling of the NPU primary circuit;

- CW and acceptance testing for the SRS and SS systems;

- physical start-up and measurement of the neutron and physical characteristics of the reactor;

- performance of IHAT and thermotechnical measurements.

Brief characteristics and scope of the works shall be provided for each CW and testing stage and sub-stage; peculiarities and purpose of the stages (sub-stages) shall be also reflected, the ways to perform the works at the RP (including safety systems) shall be specified, and the interface with any other constructed or operating RPs shall be demonstrated. It should be demonstrated that the provided scopes of works at individual stages and the RP commissioning works in general are sufficient and comply with the requirements specified in par. 13.1.1.

13.1.4. Testing programs

Brief summary of the programs for each NPU commissioning stage (sub-stage) and information on the programs for individual equipment, systems and components at each stage shall be provided.

In this case the following shall be specified for each stage (sub-stage):

- purpose of the works and tests, successful completion criteria;

- sequence of the works;

- requirements for readiness of the rooms, systems and equipment for the works;

- process restrictions, limits, conditions and arrangements for safe performance of the works and tests;

- composition, sequence, interrelation and duration of tests;

- basic provisions of the work methodologies with indication of the acceptance criteria;

- requirements for the reporting documentation, particularly for its issuance, presentation and storage, as well as access to it;

- requirements for the size and qualification of the personnel engaged in the works and tests (including the administrative personnel), distribution of liabilities and responsibilities.

The quantitative and qualitative parameters of the NPU commissioning program shall be compared with the counterparts with regard to the scope, means, methodologies, methods for arrangement of the works and tests.

The stage, way and scope of the trial runs for the design steady, transient and emergency modes (the list of planned programs and works shall be provided) as well as methods and devices for operability checks of the SRS and SS systems shall be specified; the design modes not subject to any checking shall be specified, and permissibility of this testing non-performance shall be substantiated. Specific and detailed information shall be provided to confirm that the planned works and tests will enable to comply with the above-mentioned safety conditions.

The following shall be presented in details:

- the program of the reactor bringing into the critical state, measurement of the neutron and physical characteristics of the nuclear core and reactivity compensation devices and thermotechnical measurements;

- methods for assessment of the most important characteristics of the RP, SRS, SS equipment and the NPU characteristics.

The procedure for development and approval of the NPU commissioning programs, quality assurance programs in the course of commissioning and working programs based on the design documentation shall be specified.

13.1.5. Schedule of works and testing

The master schedule of the NPU commissioning works with due regard for the nuclear core loading and the NPU commissioning time limits shall be provided.

The estimated duration of the works, the list of all types of works and tests for each stage shall be presented in the master schedule. The information on the NPU with its auxiliary systems and safety systems, as well as on the steam power and electrical equipment of the NPU shall be presented separately. The planned adjustment and testing schedules shall be provided for individual ship service systems supporting the NPU operation.

The schedules shall take into account the time for performance of the works as well as for

processing, analysis and presentation of the results and their approval by the concerned parties in accordance with the established procedure. The time necessary to develop more detailed or adjusted process operations or works on the ship and approve them prior to adoption, the time for development of testing guidelines, emergency protection and operation manuals and training of the personnel shall be taken into consideration.

13.1.6. Additional requirements for the NPU commissioning

The requirements to be taken into consideration in preparation for the testing and in the course of testing on the ship shall be specified, particularly the requirements for:

- conditions for preparation, agreement and approval of the detailed design documentation for the NPU, the NPU SAR, the set of guidelines, particularly for actions in emergency conditions, etc.;

- participation of the operating and additional personnel in performance of the works and tests and development of the documentation, particularly the reporting one (including the requirements for the format of the reporting documentation and its submittal to the concerned parties);

- administrative and technical measures and actions in case of any beyond-design characteristics or deviations from the design, particularly the necessity to adjust the design and operation documentation;

- investigation of any malfunctions and accidents at the ship NPU;

- arrangement of maintenance and document management on the ship;

- arrangement of the restricted access areas in the ship rooms and security zones depending on the phases and stages of the NPU commissioning program;

- arrangement of the fire protection and control in the NPU rooms;

- arrangement of the controlled and supervised areas, radiological and radiometric control in the ship rooms and near it;

- development and issuance of the NPU sanitary certificate and data sheet;

- development and implementation of the emergency response plans for protection of the personnel and the public in case of any accident on the ship.

13.2. Requirements for the information introduced to the SAR at the NPU handover to operation

This subsection shall be developed on the basis of the SAR for the design stage supplemented subsequent to the results of adjustment works and tests at different NPU commissioning stages, including IHAT, sea acceptance testing as well as recommendations of the ship acceptance commission.

Fulfillment of the requirements introduced to the SAR during development of the works as well as compliance of the NPU characteristics, systems and components of the ship service systems with the design and regulatory documents shall be confirmed by documents based on the reporting materials and the results of works and testing.

In case of any deviations from the design or the regulatory documents the design documentation shall be adjusted by introduction of changes and additions to the relevant SAR sections with substantiation of permissibility for any deviations with regard to the required safety and reliability level assurance.

13.2.1. Organization

Any changes of the operating organization structure introduced in the course of the NPU commissioning works shall be presented. Modifications of the organization and the testing program and their reasons shall be specified.

13.2.2. Stages of works

Information on correctness of the NPU commissioning division into stages and sub-stages with due regard for peculiarities of the particular NPU and the tasks to be solved at each stage, sufficiency of the planned works at each stage, as well as information on completeness of work performance at individual stages, optimality of the sequence of performance and (or) combination of tests proposed in the SAR for the design stage, the results of their implementation control and achievement of the acceptance criteria shall be provided.

13.2.3. Testing programs

It should be demonstrated that the testing programs for individual NPU equipment and the supporting ship service systems fully comply with the programs for each stage of the NPU commissioning as well as the testing programs for each ship commissioning stage, i.e. the tests are performed within the scope prescribed in the programs, and the established results are achieved. Testing results according to the programs and their approval by the concerned parties in compliance with the established procedure shall be specified.

The planned adjustment works and testing shall be analyzed, and specific information on compliance with the safety requirements shall be provided.

13.2.4. Schedule of works and testing

Compliance with the integrated work schedule of the NPU commissioning program shall be demonstrated and analyzed from the viewpoint of completeness and time limits; feasibility of any deviations shall be assessed.

Resolution of any issues related to coordination of the works at the commissioned NPU with any other constructed or operating energy sources of the ship shall be reflected.

13.2.5. Additional requirements for the ship NPU commissioning

Any additional requirements for the commissioning and the degree of their fulfillment adequacy shall be specified.

14. OPERATION

Information on arrangement of operation, training of the personnel and operability maintenance for the engineering features and the entire NPU while complying with the operation and safety limits and conditions shall be provided in this section.

14.1. The following information shall be provided:

- the organizational structure of the operating organization with indication of the main functions of its departments at all stages of the NPU operation and decommissioning;

- the organizational structure for the ship NPU operation management with indication of the managing positions for the departments, authorities of the managers and their responsibility for nuclear and radiation safety assurance, including the NPU operating personnel, the size (with due regard for the

reserve) and the list of job descriptions;

- the ship organizational structure for nuclear and radiation safety assurance and the NPU operation management on the ship;

- the list of departments and organizations performing any activities and rendering any services for the operating organization with indication of their names, managing positions, structural units, duties of the personnel, qualification, responsibilities of the departments.

14.2. Information on the staffing, qualification and training of the personnel shall be provided in this SAR subsection, namely:

- qualification of the managing staff of the operating organization, the personnel of the ship and any organizations engaged in performance of any activities and provision of any services;

- the personnel control system and the arrangements aimed to maintain the required qualification, including training on simulators in order to exercise any actions under normal operation conditions, in case of any operational occurrences and design basis accidents;

- description of the personnel recruiting, training, admittance and retraining system;

- analysis results for the state of training facilities and simulators, any measures to compensate absence of the full-scale simulator;

- state of the ship crew training for the ship damage control, particularly mitigation of any nuclear and radiation accidents, on the simulators, in training centers and on twin ships.

14.3. Information on maintenance and nuclear and radiation safety assurance for the ship NPU shall be presented in this SAR subsection:

- implementation of the technical regulations for the NPU and SS equipment;

- development of the maintenance and repair plans including the annual plan and the preventive maintenance schedules with indication of the basic types and scope of the works in accordance with the operational guidelines;

- efficient and timely assistance of the design organizations;

- implementation of the technical regulations for the equipment and systems ensuring physical protection of the NPU;

- planned arrangements for nuclear and radiation accident management, conditions of the emergency response control posts;

- surveillance over safety assurance in the course of the NPU operation by analysis of compliance with the requirements of the operation documentation and its maintenance;

- planned works of the RP, NPU, IHCS designers;

- peculiarities of the technical availability maintenance for the SS systems and components in case of long-term storage and preservation.

14.4. The following information shall be provided in the SAR:

- arrangement of the watch and duty sections and monitoring of the NPU;

- the procedure for maintenance of in-process records on the NPU operation, their documentation

as well as the radiation situation, exposure doses and all events related to the NPU operation, the procedure for the information storage and presentation;

- the list of potentially nuclear-hazardous works and technical requirements for their performance in the course of the NPU operation;

- maintenance of the operation documentation and guidelines;

- availability of any guidelines on maintenance of the operational documents;

- the procedure for classification, investigation and presentation of information on any ship NPU operational occurrences;

- the state of institutional control over the operation level, the procedure for presentation of information on the ship NPU safety;

- the NPU operation data and parameters included into the annual report of the operating organization on safe NPU operation.

14.5. Information on the ship NPU operation defined in the Quality assurance program for the operation stage shall be provided in the SAR.

15. ANALYSIS OF ACCIDENTS

The ship NPU safety assessment in the SAR shall include analysis of the reactions of the NPU systems and the entire ship on any potential initiating events. The analysis shall be performed in order to determine the sequence of events (scenarios) and conditions of their development with due regard for any dependent and independent failures and damages of the systems and components or human errors aggravating the situation.

This analysis shall constitute an integral part of the ship NPU safety analysis.

The scenarios of the expected events and their consequences as well as the possibility for intervention into the operation of the systems in order to control the progress of the processes shall be defined in this section of the SAR.

This analysis shall be used as the basis for arrangement of control for the NPU systems in different situations.

In the course of the analysis independent failures, undetected failures, any external common cause failures related to the ship accident and human errors shall superimpose each expected initiating event.

Safety analysis shall be performed in accordance with the lists of initiating events used as the basis to develop the lists of design basis and beyond design basis accidents.

15.1. The list of initiating events

15.1.1. Classification of initiating events

The recommended list of initiating events is given in Appendix to this section. This list shall be adjusted and substantiated for the ship NPU under consideration.

Each initiating event shall be analyzed in combination with various failures and any other factors in order to select the most significant scenarios for the analysis.

Initiating events shall be grouped into classes in accordance with their functional impact on the RP:

a) internal:

- increase of heat removal from the primary circuit;

- decrease of heat removal from the primary circuit;

- the coolant flow rate decrease;

- any changes of reactivity and power density distribution;

- the primary circuit coolant mass increase;

- the primary circuit coolant mass reduction (including loss);

- releases of radioactive media from the systems and equipment;

- loss of the secondary circuit working medium;

- loss of power supply sources;

- malfunctions in the process operations;

- spurious actuation of the systems;

b) external related to the ship accident:

- shock impacts due to collision or stranding;

- constant or periodic loss of the cooling water supply resulting from the ship stranding (probably with heeling);

- the ship capsizing;

- the ship drowning in shallow and deep water;

- crash of a helicopter with the weight of 10 t onto the NPU rooms from the height of 50 m;

- fire or explosion on the ship.

15.1.2. Causes and identification of the initiating events

The particular initiating events shall be defined for each class of initiating events, and their causes shall be considered. Larger amount of information shall be provided for the events resulting in more severe consequences (for example, all possible sequences of the accident events shall be analyzed with due regard for quantitative parameters of their occurrence probability).

In case the event does not result in any hazardous consequences according to the expert assessment qualitative description of the potential consequences shall be sufficient.

Expert assessment of any qualitative changes in the main parameters during the initiating event that may be used for identification of the initiating event shall be performed.

15.1.3. Analysis of any potential development paths (scenarios) for the emergency situations associated with the initiating events and compilation of the list of design basis accidents

The following shall be specified for each event:

- activation sequence for mechanisms and systems, generation of signals, actuation of the setpoints with regard to preventive and limit (design) parameter values, the necessary actions of the personnel, etc.;

- time limits for the SS operation commencement and completion;

- impact of the functioning normal operation systems on the process development;

- assessment of information on the situation development necessary for the operating personnel including the instrumentation readings.

The SS functions used in safety assessment, any uncertainties associated with each of the above-mentioned functions, the expected and maximum delay time shall be specified.

Qualitative assessment of the initiating event consequence severity with due regard for any coinciding independent and dependent failures or erroneous actions of the personnel shall be provided within the scope determined by the standards. Sequences (chains) of events and failures that can result in the most severe consequences (the maximum pressure increase in the primary circuit, the minimum burnout ratio, the maximum exposure dose, etc.) shall be defined for the type (group) of initiating events under consideration based on the above-mentioned assessments.

Preliminary expert assessment of any potential accident sequences shall be the mandatory element of the analysis used to compile the list of design basis accidents for further quantitative analysis according to the maximum severity of consequences but without any exceedance of exposure doses and release norms.

15.2. List of beyond design basis accidents

15.2.1. Scenarios of beyond design basis accidents resulting in excessive releases of radioactive substances into the environment Vulnerable points of the NPU

All scenarios of beyond design basis accidents resulting in exceedance of exposure doses for the personnel and the public and the norms of RSb releases and content in the environment established for design basis accidents shall be specified based on the analysis results according to par. 15.1.3. Vulnerable points of the NPU shall be defined through the minimum cut sets of event (failure) trees (hereinafter the term "vulnerable points" shall mean combinations of the NPU design peculiarities, its schematic solutions, layout, operational procedures and organizational structure of the personnel's activities representing the most probable reasons of the reactor nuclear core damage in excess of the permissible damage limits for design basis accidents).

15.2.2. Characteristic groups of beyond design basis accident scenarios

The scenarios specified in par. 15.1.3 shall be united into groups; the response required to prevent the accident development shall be the same within each group (the system functional event trees shall be similar).

15.2.3. Representative scenarios of beyond design basis accidents

One or several representative scenarios complying with the following four criteria in total shall be defined within each group in par. 15.2.2:

- the maximum dose rates for the personnel and (or) the public;

- the maximum intensity of radioactive substance releases;

- the maximum integral RSb release;

- the maximum (most hazardous) damage for the NPU and ship equipment and systems.

15.2.4. List of beyond design basis accidents

The scenarios defined in par. 15.2.3 shall be summarized in the list of beyond design basis accidents for further analysis.

15.3. Analysis methodologies

15.3.1. The list of applied methodologies

The list of methodologies used for the quantitative analyses shall be provided with indication of the data on their validation in the Software Validation Board of the Russian Gosatomnadzor. The number of the certificate, date of issuance and the certificate validity period shall be specified. In case any calculation methodology was not submitted for validation the planned time limits for validation shall be specified.

Information provided for the analysis methodologies and the time limits for expert review of each methodology depend on availability of the certificate for this software tool.

15.3.2. Description of mathematical models

The model of the analyzed processes shall be described. The main physical phenomena determining the process development shall be listed.

The system of master equations in the mathematical model shall be presented in the form it was converted into from the canonical form of record for direct usage in this calculation model. Constitutive relationships shall be provided, and the applied nodalization scheme and the numerical computation method shall be described.

Mathematical models describing transfer of fission products in the nuclear core, the NPU circuits and systems shall take into account physical and chemical processes affecting any changes of RSb concentration in the NPU circuits and rooms subject to ingress of radioactive substances in the course of the accident scenario under consideration. The minimum set of these processes shall be the following:

- natural deposition on the inner surfaces;

- desorption from the inner surfaces to the steam-gas medium;

- radioactive decay;

- leakage through any sealing defects into the adjacent rooms and the environment together with the steam-gas medium due to pressure differential;

- leakage to the environment after pressure equalization due to free convection determined by the temperature difference and media composition in the room and the atmosphere;

- purification of the steam-air medium flowing through passive condensing devices (bubblers);

- purification of the steam-air medium due to operation of the spray system;

- purification of the steam-air medium due to operation of the special ventilation system;

- chemical reactions in water changing physical and chemical properties of the fission products;

- chemical reactions in the steam-gas phase and on the surfaces changing physical and chemical properties of the fission products;

- water purification from radioactive products. The mathematical models shall take into account behavior of aerosol particles and fission products united into groups in accordance with their physical and chemical properties. The groups under consideration shall include:

- inert radioactive gases;

- volatile (organic and inorganic) iodine forms.

The mathematical models shall contain only substantiated values of the coefficients characterizing the modelled physical processes (diffusion, desorption, clearance, etc.). In case any new coefficients are used their application shall be substantiated, and credibility of the applied values shall be demonstrated.

The applied mathematical models shall contain substantiated values of the weight proportion adopted in the calculations for radioactive iodine in molecular form, in the form of organic compounds in in aerosol form.

The information shall be illustrated with the required graphic materials (schemes, block diagrams, graphs) in order to explain interaction of the programs and information transmission from program to program, particularly in case of necessity to adjust the calculations due to changes in the input data.

In case any particular processes are not considered in the models conservatism of the assessments shall be demonstrated.

15.3.3. Assumptions and calculation method errors

All assumptions and simplifications used in the mathematical model shall be specified. Permissibility of these simplifications shall be substantiated. Conservatism introduced by the adopted assumptions and the methodology error shall be assessed.

15.3.4. Application scope for the calculation methodologies

The application scope shall be defined the applied calculation methodology stated or planned to be stated in the validation certificate. The boundaries of the application scope shall be based on the relevant verification results. The possibility to use the calculation methodology for the performed analysis shall be substantiated.

15.3.5. Information on verification of the calculation programs

Mathematical models of emergency modes used for safety analysis, developed accident management programs and mathematical support of the simulators shall be compared with the experience data. Verification matrix shall include all experimental facilities used to substantiate the software tools.

Completeness of verification information shall be defined by availability or absence of the validation certificate. In case the certificate is available only references to the relevant registration number and the verification report shall be given, and in absence of the certificate information on the experimental facilities, standard problems and processes for which verification calculations were performed in accordance with the program, status of these calculations (post- or pre-testing, etc.) and description of the obtained results shall be provided. This information may be provided in a separate comparative report attached to the SAR.

15.4. Input data for the calculations

The list of input parameters and initial conditions enabling to perform repeated calculations in case of necessity shall be provided.

15.4.1. Engineering input data

The main structural characteristics (volume, length, passage areas, elevation differences, heat exchange surfaces, weights, thickness of partitions, hydraulic diameters, local resistance) shall be specified for:

- the reactor, the nuclear core, the main circulation circuit, the SG, the pressurizer;

- the steam pipelines, the ECCS hydroaccumulators, the system of leak-tight containment rooms.

15.4.2. Physical input data

The following shall be specified:

- neutron and physical characteristics (variation and reactivity coefficients, differential and integral efficiency of the CPS, prompt neutron lifetime, share of delayed neutrons, etc.);

- thermal and physical characteristics (thermal conductivity, heat capacity and density of the applied materials; temperature and enthalpy of different make-up sources and storage tanks; position of phase levels and mass in vessels with phase separation);

- physical and chemical properties of reagents and solutions generated in the course of accident, their radiation stability, partition constants and chemical reactions with basic iodine compounds.

15.4.3. Technological input data

The design characteristics (operation algorithms, setpoints, characteristic parameters, characteristics of the main equipment - pumps, discharge devices, heaters, etc.) shall be provided for the following systems:

- EP; systems for: pressure maintenance in the primary circuit; pressure maintenance in the secondary circuit; feedwater; steam removal; ECCS; PHRS; protections and interlocks; spray system; hydrogen after-burning; ventilation; containment drainage; the containment; characteristics of the pumps (the PCP, the main and emergency feed pumps, the ECCS pumps, spray pumps); characteristics of the valves.

15.4.4. Topological input data

In case any calculation schemes (nodalization schemes) are used the relations between the calculation elements and connections shall be illustrated with indication of elevations and peculiar points (points of leakage, make-up, valves, etc.).

15.4.5. Initial conditions

The list of ibitial conditions shall be provided. They should be conservative for the analyzed process. The conservatism degree shall be assessed.

15.5. Analysis of design basis accidents

15.5.1. Description of the sequence of events and operation of the systems

Description of the sequence of events and operation of the systems based on the analysis results shall be provided in the form of the table including characteristic points for the process with indication of the relevant time moment.

15.5.2. Safety assessment criteria

Based on the fact that the parameters determinative for safety can go beyond the permissible limits in the modelled emergency mode the relevant criteria shall be specified; comparison of the obtained results with these criteria will enable to assess safety of the facility under consideration in this emergency mode.

15.5.3. Analysis of the calculation results

Information shall be provided for all stages of the transient process or accident. Transition to the steady mode with functioning in accordance with the design scheme for normal operation or to steady functioning of at least one SS channel with the cold equipment parameters may serve as the sign of process completion.

15.5.3.1. Changes of parameters in the NPU circuits

The following information shall be provided:

- power changes;

- pressure changes in the circuits;

- temperature changes for the coolant, fuel element claddings and fuel;

- sub-critical thermal flux safety margins;

- coolant flow rates in the reactor and the loops;

- parameters of the primary circuit coolant at the inlet and outlet and in the most stressed channels;

- thermotechnical characteristics of the nuclear fuel;

- parameters of the secondary circuit coolant;

- hydrogen releases from the primary circuit;

- flow rate and enthalpy of the coolant discharged from the circuit;

- hydrogen amount in the primary circuit, results of comparison of the design values with permissible ones.

15.5.3.2. Changes of parameters in the containment rooms

Processes developing in the containment rooms shall be described in detail. At least the following shall be specified:

- pressure in leak-tight rooms;

- characteristics of the existing leaks from the systems to the rooms (leakage flow rates, flows through discharge and safety valves, temperature);

- characteristics of leakages to the environment (flow rates, total released weight);

- characteristics of hydrogen sources;

- characteristics of the spray system operation;

- operation characteristics of the system for heat removal from the protection rooms;

- media temperatures in the atmosphere of protection rooms and floorings;

- water and steam mass in the atmosphere of the rooms and water mass on the flooring;

- temperature of partitions and structural elements;

- relative shares of the components in the atmosphere of protection rooms, including hydrogen.

15.5.3.3. Release and propagation of radioactive substances

The adopted assumptions, parameters and calculation methods used to determine the exposure doses resulting from accidents shall be described in this paragraph.

Fission product transfer processes in the containment rooms shall be described in detail.

At least the following shall be specified:

- accumulation of the fission products in the nuclear fuel as of the accident moment;

- thermal and physical characteristics of the atmosphere and internal surfaces of the process rooms along the fission product migration routes;

- time-dependent leakage of fission products from under the fuel element claddings and from the primary circuit;

- characteristics of the main processes for transfer and deposition of the fission products in the NPU process rooms with due regard for changes of phases, physical and chemical form as well as leakage of the fission products into the environment.

This paragraph shall contain all necessary input data enabling to perform independent analysis of:

- the calculation parameters;

- places and areas where the doses are calculated, including the ship rooms (the CCR, the ECS, safety systems, rooms where the installed equipment shall be monitored and maintained, cabins, boundaries of the estimated areas).

Reference to any consolidated or validated programs used in the design may be given.

In case no radioactive substances are present beyond the boundaries of any barrier the value (or parameter) of the available margin, reserve, etc. ensuring confinement of radioactive substances within the established boundaries shall be characterized.

Results of the analyses shall be presented in the form of tables.

In case it is impossible to include any results into the table due to large amount of materials they may be presented in a separate paragraph, or reference to the relevant materials containing the required information in sufficient detail shall be given.

This paragraph shall contain the detailed dose rate calculation scheme for any damage of the protective barriers including leakages from the containment (the leak-tight circuit) with the relevant explanations for the adopted model. All possible paths of activity leakage and transfer from any room to

other rooms and the environment shall be analyzed in the scheme, the safety assurance means (filters, sprayers, membranes, partitions, etc.) and the medium flow direction shall be specified.

Several schemes for various period or cases may be provided.

Special attention in analysis of the applied assumptions and methodologies for assessment of radiological consequences shall be paid to the fact that they should be properly confirmed by the accumulated data through description of the relevant information with the reference to other SAR sections or regulatory documents. This information shall include:

- description of the applied mathematical or physical models, particularly simplifications and approximations;

- definition and description of the computer codes or analog systems used in the analysis. Description of the applied mathematical models and programs shall be arranged by brief summary of their contents in the SAR text or giving references to the sources;

- determination of time-dependent characteristics, activity and leakage rate for the fission products or any other radioactive substances carried over in the containment system that can ingress into the environment via the leakages through the containment boundaries and the ventilation system;

- consideration of any uncertainties with regard to calculation methods, equipment characteristics, sensitivity of the instruments or any other uncertainties taken into account in the result assessment;

- degree of interconnection between the systems directly or indirectly affecting control and mitigation of leakage from the containment system or any other sources (for example, from the NPU drain tanks). For example, the contribution of the following systems: spray, ventilation and conditioning, reactor cooldown and purification, radiological control, etc.

Results with regard to the absorbed doses for the child's thyroid and the external exposure doses at the boundary of the sanitary-protective area (in case of any accident in the port or basing area), absorbed doses in the ship rooms for various periods with indication of characteristic stages (overpressure existence time, destruction time, activation of devices, duration of watch or shift) shall be provided in this paragraph. Information for the operating personnel shall be presented separately. The accident development phases shall be characterized; areas of potential radioactive contamination (pollution) shall be described on the basis of the calculation data with regard to equivalent dose rate, equivalent dose of external and internal exposure for the public due to inhalation of radioactive aerosols at various distances from the accident place.

Depending on the accident type and consequences the scope and particularization degree for the provided information shall increase in proportion to the accident severity.

15.5.4. Conclusion

Conclusions on the main analysis results including definition of the most severe modes and grounds for conclusions with regard to the ship safety under design basis accident conditions shall be made.

15.6. Analysis of beyond design basis accidents Development of arrangementsfor beyond design basis accident management

15.6.1. Description of the sequence of events, operation (failures) of the systems in case of beyond design basis accidents

Sequence of events, activation, failures of the systems (components) shall be described for beyond design basis accident scenarios. It is recommended to present development of the accident events in the form of the table containing the main stages and the corresponding time moments.

15.6.2. Results of the calculation analysis

15.6.2.1. Changes of thermal and hydraulic parameters in the NPU circuits

Thermal and hydraulic processes taking place in the primary and secondary circuit of the NPU shall be described for all beyond design basis accidents from the compiled list. The scope of the presented information shall cover at least the following parameters and initial conditions:

- the reactor power;

- characteristics of heat fluxes;

- pressure changes in the circuits in the course of the emergency transient process;

- temperature changes in the coolant and fuel elements in the nuclear core;

- coolant flow rates in the reactor;

- parameters of the primary circuit coolant at the inlet and outlet of the most heat-stressed channels of the nuclear core;

- thermotechnical characteristics of the fuel;

- parameters of the secondary circuit working medium;

- coolant flow rate in different systems affecting development of the emergency transient process;

- weight of zirconium (if any) having reacted with water steam in the nuclear core;

- flow rate and enthalpy of the coolant discharged from the circuit.

15.6.2.2. Changes of parameters in the containment rooms

Thermal and hydraulic processes taking place in the containment rooms shall be described for beyond design basis accidents accompanied with release of the coolant and (or) nuclear core materials from the primary circuit to the containment. The scope of the presented information shall cover at least the following parameters:

- pressure in the rooms;

- characteristics of heat fluxes.

15.6.2.3. Interaction of molten fuel with the reactor and the containment structures

Thermal and hydraulic processes taking place in the reactor caisson shall be described for beyond design basis accidents accompanied with melting and release of the nuclear core materials from the reactor pressure vessel to the containment. The scope of the presented information shall cover the following parameters:

- changes in the state of corium components;

- corium temperature changes;

- characteristics of heat fluxes;

- the caisson configuration change;

- weight (share) of zirconium and other metals having reacted with water steam;

- characteristics of steam explosions (released energy, parameters of shock waves affecting the reactor vessel and other NPU and containment structures).

15.6.2.4. Release and propagation of radioactive substances

Fission product transfer processes in the containment rooms shall be described in detail with indication of the following information:

- accumulation of the fission products in the fuel as of the accident moment;

- thermal and physical characteristics of the atmosphere and internal surfaces of the process rooms along the fission product migration routes;

- time-dependent leakage of the fission products from heated and melting fuel and the primary circuit;

- characteristics of the main processes for transfer and deposition of the fission products in the NPU circuits and rooms with due regard for changes of phases, physical and chemical form as well as leakage of the fission products into the environment.

15.6.3. Beyond design basis accident management arrangements

15.6.3.1. Immediate safety objectives

The immediate safety objectives shall be defined for each beyond design basis accident severity level, i.e. the objectives the ship operating personnel shall strive to achieve under these conditions in order to prevent or stop further development of the equipment and (or) SRS damage or to limit releases of radioactive substances into the environment.

15.6.3.2. Facility condition characteristics, criteria of the beyond design basis accident occurrence and development

The facility condition characteristics shall be defined based on the performed calculation analyses of beyond design basis accidents, and the criteria that can be used (together with the condition characteristics) to determine the fact of the beyond design basis accident occurrence and to trace development of the relevant accident severity levels shall be established.

15.6.3.3. Systems and equipment that can be used to achieve the safety objectives and to mitigate the accident consequences

All technical systems of the ship (including the systems not related to safety assurance) that may be used (perhaps for the purposes other than the design ones and not in the design operation modes) in order to achieve the immediate safety objectives and to mitigate the accident consequences at each severity levels shall be defined. Issues of redundancy shall be considered for the systems performing the same function.

15.6.3.4. Success criteria for corrective measures

Success criteria for the actions of the personnel aimed to achieve the immediate safety objectives at each accident severity level shall be defined. These criteria shall be determined through the condition characteristics.

15.6.3.5. Analysis of the scope of information on the facility condition available to the operating

personnel in the course of accident development

The scope of information required to monitor the facility condition characteristics, to establish the accident severity levels, to control the necessary technical systems and to assess success of any actions for beyond design basis accident management as well as any hardware and techniques enabling to obtain this information under the expected conditions shall be defined. In case of necessity to perform indirect assessment of the required parameters the methods of such assessment shall be provided.

15.6.3.6. Strategy of corrective actions

The strategy of corrective actions of the personnel in case of any beyond design basis accident aimed to achieve the safety objectives at all potential accident severity levels shall be described.

15.6.4. Efficiency assessment for the proposed beyond design basis accident management arrangements

It should be demonstrated by calculations that implementation of the planned strategy of corrective actions in case of a beyond design basis accident caused by manifestation of any detected vulnerability at all possible accident severity levels will ensure interruption of the accident process development or significant mitigation of the accident consequences.

15.6.5. Conclusion

Conclusions on efficiency of the developed beyond design basis accident management arrangements shall be made based on the materials provided in Subsection 15.6.

Appendix to Section 15

LIST OF INITIATING EVENTS

(RECOMMENDED MINIMAL LIST)

INTERNAL EVENTS

1. Increase of heat removal from the primary circuit

1.1. Malfunctions in the feedwater flow rate control system with decrease of feedwater temperature

1.2. Malfunctions in the feedwater system with increase of feedwater flow rate

1.3. Malfunctions in the turbine set control system resulting in steam extraction increase

1.4. Activation of relief and (or) safety devices on the steam pipeline for various causes with due regard for potential valve failure to seat

1.5. Ruptures of the main steam pipeline in various points and rooms

2. Decrease of heat removal from the primary circuit

2.1. Malfunctions in the control system with the steam flow rate reduction

2.2. Pressure increase in the steam pipeline

2.3. Closure of the turbine set stop valve

2.4. Loss of steam extraction from the NPU

2.5. Loss of vacuum in the condenser

2.6. Isolation of the main condenser, loss of cooling water supply to the main condenser

2.7. Shutdown of the condensate and feed pumps

2.8. Feedwater pipeline ruptures

3. Decrease of the primary circuit coolant flow rate

3.1. Disabling of different number of PCPs

3.2. Switching of all or some PCPs to different speed

3.3. PCP blocking

4. Unauthorized reactivity change

4.1. Unauthorized movement of the most efficient reactivity control device with the speed permitted by the control system and the drive design at different power levels

4.2. Injection of cold feedwater into the SG and the reactor, malfunctions of the feed valve

4.3. Unauthorized activation of the emergency cooldown system

4.4. Installation of a fuel assembly inconsistent with the fueling pattern

5. The primary circuit coolant mass increase

5.1. Unauthorized pump start-up

6. Loss of the primary circuit integrity

6.1. Rupture of the primary circuit coolant pipeline:

- small leakage of the primary circuit pipeline;

- circuit-to-circuit leaks in the primary circuit equipment;

- the reactor cover leakage.

7. Releases of radioactive media from the systems and equipment

7.1. Equipment leakage through sealings

7.2. Leakage of pipelines in the RW transportaion and storage systems

7.3. Leakage from any tanks containing radioactive substances

7.4. Releases of radioactive substances during any NF accidents in case of:

- refueling;

- falling of SFA containers;

- SFA break-off in the medium section of the nuclear core in the course of its removal.

8. Malfunctions in the power supply system

8.1. Loss of power supply on one hull side

8.2. Failure of the turbine generator on one hull side

8.3. Loss of the CPS power supply

8.4. Loss of power supply for the actuator of one CPS device

8.5. Short-term or long-term total blackout of the NPU

9. Incidents in the course of NF handling

9.1. Falling of individual casks with fuel assemblies, casings with spent fuel assemblies, transport packages in the course of handling operations

9.2. Falling of any objects that can change the position or break the integrity of fuel assemblies and fuel element claddings

10. Violations of habitability conditions in the CCR and the NPU rooms

10.1. Malfunctions in the ventilation and conditioning system

EXTERNAL EVENTS

1. Ship accidents:

- stranding;

- collision with other ship (pier) with water ingress into the power and auxiliary compartments;

- capsizing;

- drowning in shallow water;

- drowning in deep water.

2. Shock waves due to:

- explosions on board;

- human activities in the course of the ship stay in the port;

- fire in the CCR and the power compartment, particularly in the engine room, the electrical compartment, the reactor compartment and the IHCS equipment rooms.

3. Helicopter crash:

- on the NPU rooms;

- on the ship hull structures containing potentially hazardous equipment (pressurized equipment, equipment filled with hydrogen, oxygen, aviation fuel).

4. Loss of cooling water

16. SAFE OPERATION LIMITS AND CONDITIONS OPERATION LIMITS

Information on safe operation limits and conditions and operation limits established in the design for safety systems and safety-related systems as well as for the ship NPU shall be provided in this SAR section.

The operation limits, safe operation limits and conditions shall be based on safety analysis for the ship NPU in accordance with the provisions specified in the design.

Substantiation of the safe operation limits and conditions may be accompanied with description of the calculation programs with indication of their validation data and (or) the relevant experimental studies (references to the SAR sections containing the required information may be given).

Information presented in this section and information contained in the operation documentation shall be adequate at any SAR development stage.

Supplements with the relevant safety state substantiation may be introduced to this section in the SAR for the commissioning stage.

16.1. Safe operation limits

16.1.1. SS actuvation setpoints

All SS activation setpoints shall be specified, the adopted setpoint values shall be substantiated, any modes that determine reaching of these values as well as accuracy and principle for generation of the SS activation commands shall be provided. Activation setpoint values shall be specified for the warning and emergency alarms with substantiation of the interval between these setpoint values.

16.1.2. List of controlled parameters and safe operation limits for the NPU

All controlled parameters, exact points of their measurement, substantiation of the adopted value and its measurement accuracy, the measurement method, the measurement ranges, accuracy of the performed calculation and (or) experimental substantiation of the parameter, the permissible information loss period, redundancy of the measurement channels shall be provided.

The limits of controlled parameters any deviation from which represents violation of the safe operation limits and results in an operational occurrence or accident shall be specified.

16.2. Operation limits

16.2.1. Limit values of the process parameters

The selected parameter values in the operation modes, accuracy of their measurement,

measurement points, redundancy of measuring channels, permissible time of information loss shall be substantiated.

16.2.2. Process protections, interlocks and automatic controllers with their actuation setpoints

Values of the process parameters for activation of the process protections, interlocks and automatic controllers shall be substantiated. The adopted values of the process parameters shall be substantiated for the permissible modes. Location of the sensors, their redundancy, power supply, setpoint values for activation of warning alarms shall be specified; the intervals between activation of the warning alarms, actuation of the process protections and interlocks and SS activation setpoints shall be substantiated.

16.3. Safe operation conditions

16.3.1. Power levels and permissible normal operation modes

The permissible normal operation modes, for example operation at partial power levels, operation with the PCPs shut down, warm-up and cooldown modes, etc. and the corresponding power levels shall be specified. Definitions of the above-mentioned modes shall be provided.

The operation limits of the parameters such as power, pressure in the primary circuit, level in the pressurizers, temperature, radioactivity of the primary and secondary circuit coolant shall be provided for the permissible normal operation modes and each power level (reference to Subsection 16.2 may be given).

The avove-mentioned limits shall be expressed via the parameter values controlled by the operator, otherwise the relation between the limiting parameter and directly controlled parameters shall be demonstrated through the use of relevant tables, diagrams or calculation methods.

Any imposed restrictions for the permissible power levels and the permissible normal operation modes shall be substantiated with references to the relevant sections of the SAR.

16.3.2. Safe operation conditions and configuration of operable systems and equipment required for the NPU start-up and operation in the permissible modes

Information on configuration and state of the systems that should be in operable or readiness condition for the NPU start-up and operation in the permissible modes shall be provided.

Requirements for safety systems and safety-related systems shall be specified.

Configuration and quantity of the equipment to be operable for the NPU start-up and functioning in normal operation modes shall be provided for each system.

The following requirements shall be specified:

- for leak-tightness of the systems, quantity and quality of working media;

- for activation of the equipment including the sequence of actions, operation logic of the automation devices and internal protections;

- for characteristics of the systems;

- for SSSs (power supply, cooling systems, conditioning, ventilation, etc.) and the operator's intervention conditions.

Conditions related to the permissible loading cycles for the main equipment with due regard for

the design useful life as well as substantiation of the established requirements and conditions shall be provided.

16.3.3. Permissible power levels and permissible period of the reactor power operation in case of any deviations from safe operation conditions

Information on the permissible period of the reactor power operation and the power level or the NPU state in case of any deviations from safe operation conditions shall be provided.

The way to bring the NPU into the required state shall be specified, and the selected conditions shall be substantiated.

16.3.4. Recommendations for assessment of duration and permissible NPU operation power levels beyond the design limits under extreme conditions from the viewpoint of ship safety shall be presented.

16.3.5. Conditions for maintenance, testing and repair of safety-related systems

Conditions for testing, inspections, maintenance and repair of the systems specified in par. 16.3.2 shall be specified.

Information on the time limits, scope, methods and means for performance of these works and any operation restrictions (in case of necessity) shall be provided.

16.4. Administrative conditions and documenting of informationon monitoring of safe operation limits and conditions

The requirements for the ship administration and the NPU personnel with regard to compliance with the established safe operation limits and conditions shall be provided.

The list of standard documentation shall be provided, and the procedures for registration and documenting of all deviations from the safe operation limits and conditions as well as for control of their observance shall be described.

17. QUALITY ASSURANCE

17.1. General provisions

17.1.1. Requirements for information on quality assurance for any works and services affecting the NPU safety to be submitted by the applicant within the scope of SAR in order to obtain the licenses for the NPU construction and operation as well as for any other activities related to the NPU safety shall be provided in this section.

17.1.2. The information presented by the applicant in this section of the SAR shall ensure that the design, construction and operation of the NPU will be carried out properly and comply with the specified quality assurance requirements.

17.1.3. Information on the areas of activity described in Subsection 17.2 shall be provided at the relevant licensing stage in order to assess acceptability of the applicant's quality assurance activities.

17.1.4. This section shall be divided into subsections with the titles corresponding to the relevant areas of quality assurance activities in accordance with Subsection 17.2.

Information of this section shall be prepared with due regard for the results of QAP development and implementation.

17.1.5. The regulatory documents used for development and implementation of the quality assurance measures shall be specified for each area of the quality assurance activities presented in the SAR.

17.1.6. The following shall be provided together with the SAR:

a) at the ship construction license obtaining stage:

- the general quality assurance program for the NPU - QAP (G);

- the NPU design quality assurance program - QAP (D);

- the NPU development quality assurance program - QAP (DV);

- the NPU construction quality assurance program - QAP (C);

- the NPU commissioning quality assurance program - QAP (CW);

- the list of QAPs (developed and planned for development) in the course of development and manufacturing of the safety-related NPU equipment, items and systems - QAP (DV) and QAP (M);

b) at the operation license obtaining stage - the NPU operation quality assurance program (QAP (O)).

17.2. Requirements for the information on quality assurance activity areas

17.2.1. Quality assurance policy

Quality assurance policy of the operating organization (the applicant) signed by the top official shall be provided.

It should be demonstrated that the quality assurance policy:

- is coordinated with other areas of the operating organization activities;

- defines principles and objectives adopted in order to ensure the NPU safety and foreground in relation to any other objectives;

- is communicated to all executives;

- is aimed to encourage the workers assisting in detection of any potential non-conformities affecting the NPU safety.

Information confirming implementation of the operating organization policy in the area of quality assurance shall be provided in this subsection.

17.2.2. Organizational activities

17.2.2.1. Quality system of the operating organization

The following shall be specified:

- the quality system structure of the operating organization;

- the list of the main quality system documents (quality guidelines: the general guidelines and the guidelines for individual areas of activities, etc.);

- regulatory, organizational and methodological framework of the quality system;

- responsibility of the parties for quality assurance;

- structure of the quality departments;

- authorities, responsibility, direct functional duties immediately performed by the operating organization;

- infrastructure of the operating organization formed by specialized enterprises and organizations to which it partially delegates its functional duties, authorities and responsibility retaining overall responsibility without any prejudice for liabilities and legal responsibility of contractors;

- the reporting documentation containing efficiency analysis for the quality system of the operating organization, results of its inspections and corrective measures.

17.2.2.2. Arrangement of the NPU construction

The organizational structure of the departments, functional duties and authority levels for the NPU personnel ensuring the NPU safety shall be described, and the following information shall be also provided:

- the system of internal and external communications;

- structure of the quality assurance departments;

- the general design arrangement scheme demonstrating interaction of the operating organization, the main organization for the NPU design development and their contractors as well as the design approval procedure;

- the list of documents forming the legal framework of the operating organization activities;

- the procedure for the SAR development and issuance at different licensing stages;

- information on the control system of the operating organization and communication lines between the operating organization and its contractors provided for all quality assurance works;

- the list of managerial positions with the authorities and responsibility for implementation and efficiency of the general and individual quality assurance programs.

17.2.2.3. Quality assurance programs

The following shall be specified:

- information on development, approval and results of inspections with regard to implementation of quality assurance programs (general and individual) in accordance with the RD provisions;

- information on implementation of the general and individual QAPs as of the moment of the SAR submittal by the applicant;

- information on the QAP compliance with the RD provisions;

- the QAP application scope;

- information confirming that any activity affecting safety-related NPU systems and equipment is subject to the relevant control within the QAP framework;

- description of the measures implemented prior to the SAR submittal for review (particularly

terms of reference for feasibility studies, development of the NPU, the NPU construction project, etc.);

- description of the measures taken by the operating organization in order to ensure current implementation of the QAP;

- information on analysis of the regulatory and technical support performed by the operating organization at all NPU construction and operation stages;

- description of the measures taken by the operating organization in order to ensure development of the regulatory documents according to the needs detected in the course of analysis.

17.2.3. Staffing and training of the NPU personnel

Information on presence of any requirements for qualification in the job descriptions for the personnel engaged in the works affecting the NPU safety assurance as well as the scope of skills and knowledge corresponding to the specified qualification shall be provided in this subsection.

Information on the effective procedure for the NPU personnel management shall be provided with regard to:

- training, knowledge and skill checks for the personnel engaged in any works affecting the NPU safety assurance and supervision over these works (particularly the personnel carrying out tests, inspections and checks);

- determination of the needs for training of the NPU personnel and organization of training, retraining, development of competence and certification of the personnel including issuance of the relevant certificates;

- analysis of programs for training, retraining, development of competence and certification of the NPU personnel;

- maintenance of records on training, retraining, development of competence, and certification of the NPU personnel.

17.2.4. Regulatory documents

The subsection shall contain the list of regulatory documents on quality assurance (or reference to this list) effective in the operating organization and (or) any organizations performing works and rendering services to the operating organization (for example, federal rules and regulations in the area of atomic energy use, national and industrial standards, corporate standards, effective quality system procedures).

The quality system procedures planned for development in order to ensure compliance with the requirements of the SAR and the adopted quality assurance policy shall be specified.

17.2.5. Documentation management

The subsection shall include the information on the effective procedures for development, agreement, approval, implementation, identification, introduction of changes, review, distribution, storage, disposal of ineffective documents (drawings, manuals, methodologies, data. etc.) aimed to guarantee:

- rewiew and approval of any changes introduced into the documents by the organizations having approved the initial documents;

- availability of the documents at the work site prior to commencement of the works;

- timely withdrawal of the cancelled documents.

The subsection shall include the procedure and planning for the development of any missing quality system procedures.

17.2.6. Design control

The subsection shall contain the following:

- description of the measures (procedures) planned and implemented by the operating organization within the framework of the general quality assurance program for design control that shall prescribe correctness verification for the applied solutions as well as their compliance with the design requirements;

- analysis of feasibility and subsequent implementation of the initial design requirements within the terms of reference for the NPU design, development of the NPU and equipment (in this case attention shall be paid to the requirements for the NPU safety and reliability);

- description of the design verification methods used by the operating organization with substantiation of their application (for example, the design review through the use of alternative calculations or testing);

- verification of compliance with the requirements for documenting of the inspection results so that to enable investigation or revision of the inspection method after its completion;

- verification of compliance with the requirements for time limits of the inspections that shall be completed after prototype or commercial prototype validation testing prior to issuance of any documentation for manufacturing or construction;

- check of compliance with the criteria of mandatory testing provided for verification of the design, the necessity to ensure representativeness of tests and modeling of the least favorable conditions defined on the basis of the NPU safety analysis;

- description of the measures aimed to define and control delimitation of works in the course of design;

- information on the measures implemented during the design of any units in order to ensure their fabricability and reflection of any fundamental (with regard to assurance of the quality parameters for the safety-related NPU structures) process requirements and control methods in the engineering documentation;

- results of any new process trials, implementation of the process equipment, control methodologies and means performed mainly at the research and development stage;

- information on availability and implementation of the procedure for control of any change introduction to the design in the course of design and manufacturing as well as in the course of the NPU operation.

17.2.7. Procurement management for equipment, components and materials as well as provided services

Information on the following effective procedures shall be provided:

- arrangement of procurement of equipment, components and materials, as well as provision of services including the procedure for selection of organizations performing works and rendering services to the operating organization (bidding arrangement);

- maintenance of the procurement documents for equipment, components and materials as well as for provision of services;

- inspection of the quality assurance programs of the organizations performing any works and rendering any services to the operating organization, assessment of their capability to perform the works or render the services to the operating organization;

- analysis of the contracts for procurement of equipment, components and materials as well as for provision of services.

17.2.8. Control of the purchased equipment, components and materials and provided services

Information on the following effective procedures shall be provided:

- arrangement of identification, control (particularly incoming one) and testing of the equipment, components and materials;

- assurance of traceability for the control and testing results;

- assurance of completeness for various types of control and testing;

- arrangement of storage, transportation, preservation and packing of the equipment;

- arrangement of supervision over compliance with the requirements for the provided services.

17.2.9. Industrial activities of the operating organization and any organizations performing works and rendering services to the operating organization

Information on the effective procedures for performance of the required operations aimed to control quality of the processes important for the NPU safety assurance shall be provided, including the following processes:

- mechanical processing and assembly of the equipment and units of safety-related NPU systems;

- cleanness assurance in the course of manufacturing;

- installation techniques for the quality-related equipment and assemblies;

- non-destructive control methods.

- welding, surfacing, heat treatment;

- installation and dismantling of the equipment, units and structures affecting the NPU safety;

- NF refueling;

- leak-tightness control for the fuel element claddings;

- leak-tightness control for the containment;

- repair of the equipment and maintenance in the course of operation.

The following information shall be provided:

- development of the list of safety-related NPU systems (components);

- presence of any requirements for quality of safety-related NPU systems (components) and works affecting the NPU safety assurance;

- the procedure and techniques for performance and control of the works affecting the NPU safety assurance;

- application of the statistical methods (in case of necessity).

17.2.10. Inspection control

Information on the results of QAP implementation by performance of inspections shall be provided, particularly:

- the lists of inspections;

- availability of the inspection programs;

- the inspection planning schedule and its implementation;

- confirmation of independence of the inspectors from the inspected works;

- availability and implementation of the QAPs;

- the instructions on the procedure for inspections of the process control points, stages of the works after which any further works are prohibited without inspection and documented permit based on the control and inspection results.

17.2.11. Control of testing

Information on the effective procedures ensuring the complete set of test types and trial of the safety-related NPU equipment, items and systems shall be provided.

The following shall be demonstrated:

- the list of tests for the equipment and systems in order to check their operability in the course of the NPU operation;

- information on the ways to reflect the product operation model, requirements for metrological support, acceptability conditions for the testing results in the testing programs;

- methods for registration and documentation of the testing results and assessment of their acceptability;

- references to the testing reports and description of the testing results with due regard for implementation of the general and individual quality assurance programs.

17.2.12. Metrological support

The following information shall be provided:

- development and implementation of the instrumentation and equipment inspection program;

- the application scope of the verification program, availability of the lists of inspected equipment and instruments;

- availability of the provision on identification of the instrumentation and testing equipment;

- the effective procedures for:

arrangement of validation, calibration, verification and identification of the instrumentation and testing equipment;

maintenance of good operating condition and servicing of the instrumentation and testing equipment;

maintenance, accounting and storage of the validation, calibration and verification reports for the instrumentation and testing equipment.

17.2.13. Quality assurance for software and calculation methods The following shall be included:

- information on the effective quality assurance programs for software and calculation methods, particularly verification of software and calculation methods;

- the list of programs used for calculations (neutron and physical, thermohydraulic, strength, etc.), design (automated design systems) and research (automated scientific reserach systems, etc.) works with indication of the registration data;

- the procedure for arrangement and quality assurance of the calculation works;

- the procedure for updating of the technology for calculation substantiation of structures at all design stages;

- information on the competence development for the work performers;

- information on usage of any validated databases for development of programs;

- information on assimilation and implementation of alternative domestic and foreign programs;

- the procedure for training of the work performers in state-of-the-art methods for numerical solution of thermophysical and other problems;

- the software validation procedure.

17.2.14. Reliability assurance

Information on the effective procedures aimed to ensure reliability of the safety-related NPU equipment, items and systems as well as the procedure of interaction and the organizational chart for the participants of the reliability assurance works shall be provided.

17.2.15. Control of non-conformities

Information on the following effective procedures shall be presented in this subsection:

- registration of any deviations from the requirements for quality of works (services) and (or) equipment (design or manufacturing errors, defects and failures of the equipment, abnormal operation modes, human errors, etc.) and their analysis;

- prevention of usage of any products that do not comply with the established requirements (for example, the procedure for segregation, disposal, documenting, identification of such products) or acceptance of any services that do not comply with the established requirements;

- arrangement of the system for collection and processing of data on any detected non-conformities, violations, defects, their causes and the implemented corrective measures;

- detection, registration and informing of the relevant organizations on any revealed deviations in the materials, equipment and components.

Information on any registered cases of decision making on the detected non-conformities, results of their control by the quality departments and analysis of the detected deviations by the operating

organization shall be also provided.

17.2.16. Corrective measures

The subsection shall include:

- information on the effective procedures for development of corrective measures aimed to prevent repeated non-conformities, particularly subsequent to the inspection results, control of their implementation, assessment of their efficiency and documenting of these activities;

- information on the effective procedures for prevention of any potential deviations and non-conformities and control of their efficiency assurance;

- the list of basic corrective measures subsequent to the QAP implementation results as of the SAR submittal moment.

17.2.17. Quality records

The following shall be provided:

- the procedure for control of information on quality assurance in the operating organization related to the NPU quality assurance;

- availability and implementation of the procedure for recording, storage and issuance of documentation that shall be maintained in accordance with the written procedures (corporate standards, guidelines);

- information on the effective procedures for compilation and maintenance of the quality assurance documentation (establishment of the record types depending on importance, identification, collection, indexing, access, filing, storage, maintenance and destruction of the registered quality data including the results of inspections, testing, verification of the processes, analysis of the supplied equipment, components and materials);

- description of the QAP implementation reporting system which shall include the procedure for compilation of:

reports on the results of any inspections carried out with regard to application of documents, quality of the developed products, quality costs, credibility assessment, etc.;

annual reports on the product quality for the certain period;

annual reports on the results of the designer's supervision in the course of manufacturing, installation, testing and operation.

17.2.18. Inspections (audits)

Information on the effective procedures for performance and presentation of the results of any independent inspections (internal and external) of the actual quality assurance program implementation state as well as its efficiency assessment shall be provided.

18. DECOMMISSIONING

18.1. General requirements for this section of the SAR shall contain:

- the decommissioning concept;

- arrangement of the decommissioning works.

Description of the administrative and technical arrangements and radiation safety assurance means in the course of the ship NPU decommissioning shall comply with the RD requirements as of the SAR submittal moment.

18.2. Stages of the NPU and ship decommissioning

Potential decommissioning options ensuring minimum exposure for the personnel and contamination of the environment with due regard for the costs as of the design development moment shall be specified.

The following stages of the ship NPU decommissioning shall be described with the maximum possible degree of particularization:

- preparation for disposal of the ship with NPU;

- final shutdown of the reactor and arrangement of subsequent continuous monitoring of its state;

- unloading of the nuclear core, RW removal, radioactivity monitoring for the remaining large-scale and hard-to-dismantle equipment in order to hold it for reduction of activity;

- decontamination of the NPU equipment and the ship rooms. Dismantling of the equipment and its removal from the ship, disposal or keeping for reuse;

- control of the materials for further unrestricted use;

- making of the decision on the ship operation mode.

The following shall be specified in description of the general decommissioning areas and techniques in the SAR section:

- the list of the works performed by the operating organization prior to the ship handover to the special-purpose plant for disposal;

- the list of materials that may be reused without any additional processing;

- the list of materials that may be reused after additional processing;

- consolidated range of the equipment that may be reused;

- the estimated scope of processing for the ship structures and equipment (in tons) according to the selected option;

- special-purpose equipment required for the NPU decommissioning;

- the database required for the decommissioning.

The following shall be specified in the SAR:

- the scheme of waste generation in the course of decommissioning and disposal of the ship with NPU and the estimated data on their classification and total amount;

- the schemes of RW handling (separately for SRW, LRW, GRW) at the special-purpose facility during decommissioning of ships with NPUs.

18.3. Materials on decommissioning of ship NPUs presented in the SAR at the design stage

18.3.1. The future decommissioning strategy for the designed NPU shall be considered in the ship NPU decommissioning options through review of various decommissioning options with description of the possible ends states of the NPU and the ship for each option and substantiation of the most feasible variant.

The way to ensure the following at all stages of each NPU decommissioning option shall be demonstrated:

- non-exceedance of the exposure dose limits for the personnel, the public and the environment;

- generation of the minimum RW amounts (volumes);

- reduction of RSb ingress into the environment down to the minimum achievalbe level.

The following shall be specified:

- description of the main NPU decommissioning stages, each decommissioning option, including the approximate list of the basic arrangements and works aimed to prepare the NPU for decommissioning;

- the approximate list of the main planned measures for safety assurance in the course of decommissioning;

- the approximate list of systems and equipment required to perform the works for the NPU decommissioning as well as the requirements for their technical condition.

The necessary substantiation provided in the NPU design shall be specified for:

- selection of corrosion-resistant structural and vessel materials for the equipment and structures;

- selection of the structural and construction materials with the minimum activation by neutrons in the course of further NPU operation;

- the design solutions limiting the possibility for transfer and propagation of activated corrosion products in the process media;

- location of the equipment in the rooms facilitating its future dismantling and decontamination of its surfaces (external and internal) in the optimal way;

- reservation of the areas on the ship or at the ship basing site for storage of radioactive wastes and reusable materials generated in the course of the NPU decommissioning;

- the possibility to locate special-purpose equipment required for the NPU decommissioning in the NPU rooms.

Selection and substantiation of the most preferable option shall be performed with due regard for the following factors:

- potential prospects and plans for further usage of the decommissioned NPU and ship;

- expected engineering and radiation state of the NPU as of the moment of the final reactor shutdown and the possibility to predict the state of the NPU and the ship at the required moment within the entire NPU decommissioning period;

- assessment of the possible hazardous radiation impact on the personnel, the public and the environment;

- effective safety assurance standards and rules;

- assessment of the amounts, types, physical state of the radioactive wastes generated in the course of NPU decommissioning;

- availability of RW handling plants and techniques;

- availability of storage facilities for radioactive wastes generated in the course of NPU decommissioning;

- assessment of the amounts, types and physical state of the reusable materials, etc.

18.3.2. Design solutions aimed to ensure safe decommissioning of the NPU and implemented at the design and construction stage include:

- substantiation of useful life of the equipment and operability of the systems required for the NPU decommissioning not only within the design service life of the NPU but also in the course of its decommissioning, or the possibility for their replacement or service life extension;

- selection of the RW handling plants;

- segregation of radioactive wastes after decontamination and conditioning according to groups of radioactive contaminations in compliance with the RW handling standards and rules;

- the transportation scheme for delivery of RW to the RW handling complex and the storage facility for conditioned RW;

- transportation of the dismantled radioactive equipment to the RW handling complex for decontamination, fragmentation and conditioning with subsequent safe relocation for disposal;

- arrangement of the transportation schemes for disposal of the activated non-demountable equipment in the course of the NPU decommissioning;

- the maximum possible removal of the wastes accumulated in the course of operation from the RW repositories prior to deployment of the NPU decommissioning works;

- reserving of additional areas on the ship and at the potential ship basing location in order to install any facilities necessary at the NPU decommissioning stage;

- reserving of areas on the ship and at the potential ship basing location for storage of reusable materials (for restricted and unrestricted use);

- arrangement of the outgoing radiological control for all materials returned into the economic turnover.

It should be demonstrated in the SAR that the ship design provides for the following:

- arrangement of the hull structure fragments with the geometric dimensions enabling to divide the protective structure part activated by radiation to the induced activity levels (high, medium, low) as well as into the parts for restricted and unrestricted use;

- radiation protection of the process and radioactive equipment (for example, the RP iron-water protection tank) in the modular option while ensuring all strength characteristics of the protective structures;

- modular option for the protective structures providing for the possibility to divide them into the zones with different RSb contamination levels and zones without any contamination;

- application of special-purpose sealing coatings (single-, double- or three-layered) in order to reduce radionuclide contamination of the structures;

- special-purpose removable sheets in the hull structures of the rooms in order to arrange installation openings providing access to the radioactive equipment for the purpose of its dismantling.

It should be analyzed whether the capacity of the regular ventilation, water supply and air purification systems is sufficient for long-term performance of the full scope of future dismantling works, or any additional systems are required.

18.3.3. Characteristics of the equipment and hull structures

The design mass and dimensional characteristics of the main, process and auxiliary NPU equipment shall be specified and analyzed from the viewpoint of future safe NPU decommissioning:

- dimensions and weight of the reactor pressure vessel, the material it is made of, the total weight of the reactor, the nuclear core dimensions, the number of fuel assemblies in the nuclear core, the dimensions and weight of the iron-water protection tank, the reactor caisson, etc.;

- dimensions and weight of the PCP, the filter cooler, the SG, biological protection of the heat exchanger, the third and the fourth circuit, the CPS drive mechanisms, etc.;

- dimensions and weight of the circuit equipment and pipelines.

In all cases where neutron irradiation takes place in the course of the NPU operation chemical composition of the materials required to calculate neutron activation of the equipment units and components and the hull structures shall be specified.

The following shall be provided:

- the range of the process and auxiliary NPU equipment, its configuration, mass and dimensional characteristics, grades and chemical composition of steels it is made of;

- information on the suitability (for example, fragmentation) and restrictions of the main, process and auxiliary equipment for dismantling and transportation in the course of the NPU decommissioning;

- information on layout of the equipment, systems and hull structures in the NPU rooms.

Information on layout of the RP equipment, systems and hull structures shall illustrate the NPU transportation flow chart, anchoring of the equipment, systems and structures inside the NPU rooms, with indication of any zones inaccessible for dismantling works (absence of any installation and process openings with sufficient size, passages, manholes; presence of any equipment and pipelines that cannot be dismantled according to the regular repair and replacement technology).

Classification of the NPU rooms shall define the rooms referred to categories I, II, III of the controlled access area and other rooms of the supervised area.

The following information shall be presented in description of the NPU procurement, maintenance and repair systems:

- the package of the process and design documentation on standard maintenance and repair works for the NPU equipment and systems;

- the design range of the tools, jigs and fixtures, the maintenance and repair means for the NPU equipment and systems;

- the design range of the components and materials providing the NPU decommissioning

according to any option;

- the range of sites and rooms for collection, processing and storage of the generated radioactive wastes;

- the range and characteristics of any additional lifting and handling devices and utilities required to perform the NPU decommissioning works.

These data shall be coordinated with the adopted design and engineering solutions aimed to ensure safe NPU decommissioning, and the following information shall be provided:

- planned solutions in this area with regard to transportation of the dismantled equipment, systems and hull structures (including radioactive ones), storage and processing of radioactive wastes;

- the spatial and planning solutions adopted in the NPU design in order to ensure performance of dismantling operations and transportation through the use of remotely controlled means including robotic ones;

- the selected structural materials for operation under the intensive radiation conditions and aimed to reduce formation of long-lived radionuclides;

- the use of detachable and modular assemblies in the main NPU equipment and systems, detachable joints and connections of the equipment parts with different RSb contamination degree.

The following shall be specified in the subsection:

- application of easily-removable coatings and other means and arrangements in the NPU design in order to restrict propagation of radioactive contaminations and to stabilize them;

- provision of the possibility to take samples from the load-bearing metalwork in order to determine their actual mechanical properties and the remaining useful life;

- decontamination of non-reusable equipment and systems and provision of the required areas (rooms) for temporary stockpiling and storage of radioactive wastes.

18.3.4. Assessment of qualitative and quantitative composition of the radioactive substances accumulated in the main, process and auxiliary equipment and hull structures for the entire operation period

Conservative calculation estimates of radionuclide content in the equipment materials and structures due to their activation under the impact of integral radiation (as of the final reactor shutdown and in a year after the reactor shutdown) shall be provided based on the design data with regard to the range of equipment and structures, their mass and dimensional characteristics, chemical composition of the materials, etc. as well as the potential operation mode and the design service life of the NPU.

The total and volumetric specific activity data shall be presented for each equipment unit and structure subjected to neutron irradiation in the course of the NPU operation.

Information on distribution of radionuclides according to depth shall be provided for the hull structures (the reactor caisson, the biological protection tank, etc.).

Conservative assessments of radioactive wastes and reusable materials generated in the course of the NPU decommissioning (with regard to mass and volume) shall be provided on the basis of such information.

The subsection shall include the analysis results for usage of at least two possible design options in order to reduce the radionuclide content in the NPU steelwork, for example:

- replacement of alloys containing great amounts of cobalt and nickel with the alloys low in these components or alloys without these components;

- decrease of cobalt, silver, niobium and nickel content in the structural materials.

The amount of radionuclides and particle size of aerosols generated in the course of dismantling works shall be assessed based on the proposed metal cutting and breakage techniques to be used for dismantling with due regard for the information on materials and particular equipment.

Gamma radiation dose rates due to individual activated assemblies of equipment and structures shall be assessed for the industrial rooms.

18.3.5. Radiological control in the course of decommissioning

Requirements for the types and scope of radiometric, spectrometric and dosimetric control at the decommissioning stage shall be defined based on the analysis of any potential ionizing radiation sources and characteristics of aerosols. The personnel protection means and the air purification means for the local ventilation systems shall be defined.

Radiological control of the radiation situation on the ship shall be carried out within the entire period of decommissioning works through the use of the on-board radiological control system. In case of necessity the system shall be upgraded and supplemented with due regard for peculiarities of the performed works.

It should be demonstrated that the radiological control system will be operable after the reactor shutdown within the entire period of the ship disposal works and will ensure the following measurements:

- specific activity of the wastes (low-, medium- and high-level) and reusable materials (for restricted and unrestricted use);

- gamma radiation intensity in the rooms;

- gamma radiation intensity for individual units and equipment of the removable assembly, the reactor vessel, etc., fragments of the reactor equipment in the course of dismantling, segregation in accordance with the radioactivity groups and transportation;

- surface beta contamination of the equipment and rooms;

- specific volumetric activity of aerosols in the air.

It should be demonstrated that the external dosimetry system ensures monitoring of ingress to the environment for any radionuclide generated in the course of the NPU decommissioning works or any mixtures thereof.

18.4. Materials presented in the SAR for the ship NPU decommissioning safety substantiation

Subsequent to the commissioning results the following shall be provided in the SAR:

- consideration and analysis of any changes in the design solutions, selection of the materials for the equipment and structures, layout of the equipment, etc. introduced at the construction stage and significant for decommissioning;

- consideration of any new technologies for dismantling of equipment and structures, the RW handling techniques, etc. that have emerged since the NPU-powered ship construction commencement and can affect selection and substantiation of the previously adopted NPU decommissioning option;

- description of the basic and the most feasible (with regard to safety, time limits, costs, etc.) NPU decommissioning options taking into account its previous operation.

The practical operation experience for any systems important for performance of decommissioning works (ventilation, air purification systems, etc.) shall be considered in the SAR, their operability and sufficiency within the entire NPU decommissioning period shall be substantiated. The SAR shall contain all additions to the design solutions actually implemented (or not implemented) after the NPU-powered ship construction completion.