THE SAFETY ASSESSMENT OF BN-1200 DESIGN USING … · THE SAFETY ASSESSMENT OF BN-1200 DESIGN USING INPRO METHODOLOGY 1 ... Safety analysis report for power plant with fast ... Safety

Valerii Korobeinikov

State Scientific Center

Institute of Physics and Power Engineering

/ Russia

THE SAFETY ASSESSMENT OF BN-1200 DESIGN USING INPRO METHODOLOGY

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Technical Meeting to Review the Updating of the INPRO

Methodology 15-17 November 2016 Vienna, Austria

Contents

• Introduction

• Russian approach to fast reactor safety analysis

• Russian Regulatory Documents for NPP Safety Provision

• Measures to improve accident management preparedness

• Assessment of BN-1200 on criteria INPRO

• Main problems of the INPRO methodology in area of safety and ways of improving

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Introduction Why fast reactors are necessary?

Coal - 8,7%

U-238 - 86,7%

Oil - 0,8%

U-235 - 0,4%Gas - 3,4%

(Data source:

for proven resources of fossil fuel – British Petrolium «Statistic review of world energy 2005»:

oil – 9,9 billion tons, gas – 48 trillion m3, coal – 157 billion tons ;

for proven resources of Unat - Federal Subsoil Resource Use Agency data - 615 thousand tons )

Relative energy potential of natural resources of Russia.

1. Full involvement of

uranium-238 in nuclear fuel cycle

2. Increased fuel resources

3. The closure of the nuclear fuel cycle

4. Reducing the volume of spent nuclear fuel

5. Disposal of long-lived waste of nuclear power with thermal neutrons 3

Introduction two-component nuclear power system

The development of a two-component nuclear power system with VVERs and SFRs in the near-term future is based on:

Russian experience in the development and operation of the SFRs (BR-5, BOR-60, BN-350) and competences obtained in the course of BN-600 operation and BN-800 development

Continuity in the basic engineering concept in the integral-type SFR and the use of many well-tried solutions, as well as the scope of the R&D work accomplished to validate new solutions adopted in the design of the commercial BN-1200 reactor

The commercialization of the fast reactors based upon the BN-1200 project provides for the development of a complex of closed nuclear fuel cycle.

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1. Russian approach to fast reactor safety analysis

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• Russian approach to fast reactor safety analysis was formed on the basis of the large experience gained in designing and operating of Nuclear reactors and, in particular, fast neutron reactors

• This experience included in itself large sodium leaks, leading to radioactive sodium releases from the primary circuit of the reactor, as well as failures of steam generator tubes causing water and steam penetration to the secondary sodium

• International experience collected in the IAEA recommendations • Regulatory documents and reactor operation regulations are

periodically updating on the base : • Russian and worldwide experience gained in preventing and

mitigating of abnormalities and accidents • Russian and international experience in designing and safety

analysis of advanced projects (GEN-4, INPRO)

Russian approach to fast reactor safety analysis (2/3)

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• 1. Regulatory Documents for NPP Safety Provision determine the set of design and safety criteria.

• 2. One of the Main document that determined common regulatory approach and common requirements to safety analysis of fast reactors in Russia is OPB-88/97.

• It includes brief list of specific requirements regulating characteristics of various type reactors with regard to safety. It also requires that the SAFETY REPORT should be issued for each reactor. SAFETY REPORT should be developed in accordance with another regulatory document – «Special standard contents of safety analysis report».

Special standard contents of safety analysis report

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I. General requirements

1. Purpose and scope of regulatory document "Requirements for the

content of the report to substantiate the safety of nuclear power plants

with fast reactors"

2. Purpose and Scope of the report

3. The procedure for preparation of the report

4. Requirements for the content and design of the report

II. Safety analysis report for power plant with fast neutron reactors

Requirements for the introduction

1. The basis for the development of the project

2. General characteristics of the NPP

3. Stage of development: report

Special standard contents of safety analysis report (cont.)

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4. Information on the operating organization and the contractor

5. Information about the developers report

6. Information on R & D

7. Characteristics of the report

Annex 1 Typical structure of these systems within the Safety analysis report nuclear plant

Annex 2 Conditions of placing nuclear power plant

Annex 3 Results of the analysis of scenarios of initial events of natural and anthropogenic origin

Annex 4 Requirements for information about programs to ensure radiation monitoring

Annex 5 Initial data for testing calculations

Russian approach to fast reactor safety analysis (3/3)

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• 3. Large number of computer codes are involved in the process of

safety justification. In accordance with Russian regulations those

codes should be certified. The procedure of certification includes in

itself verification of the code and its expertise by the team of

independent experts (Team leader is normally the representative of

“ROSTEHNADZOR”) and expertise by Special Certification Board.

• 4. Finally it should be proved that project characteristics and reactor

behavior under normal and accidental conditions satisfies the set of

design and safety criteria.

PBYa RU AS-89. Rules on Nuclear Safety of Nuclear Power Plants

Reactor Units

OPB-88/97. General Statements on Nuclear Power Plant Safety

Provision

NRB-99. Radiation Safety Codes

OSPORB-99. Basic Sanitary Rules for Radiation Safety Provision

SP AS-03. Sanitary Rules for Nuclear Power Plant Design and

Operation

NP-032-01. NPP Siting. Basic Criteria and Requirements for Safety

Provision

2. Russian Regulatory Documents for NPP Safety

Provision

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3.Measures to improve accident management

preparedness

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Implementation of additional emergency equipment for emergency water and power supply at nuclear power plants;

Confining system reliability improvement;

Implementation of emergency and post-accident sampling;

Analysis of feasibility and expediency for implementation of the reactor pressure vessel outer cooling;

Enhancement of main control room and emergency control room protection;

Qualification of safety system components for ‘harsh’ environmental conditions;

Improvement of the emergency response interaction system;

Development and implementation of Guidelines for severe accident management;

Improvement of personnel competences and preparedness

Safety criteria applied in the safety analysis of Russian

NPPs:

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Population Radiation Safety Criteria

Personnel Radiation Safety Criteria

Design Limits of Fuel Elements Damage

Nuclear Safety Criteria

Probabilistic Safety Criteria

Safety Criteria for Radwaste Handling

Defense-in-Depth Criteria

Requirements for Safety Systems

4. Assessment of BN-1200 on criteria INPRO

• According to the INPRO methodology, an assessment in the area of safety should be primarily based on a comparison of the innovative reactor design against a reference design, i.e. one that has been put in operation by the same supplier and could be considered as its most advanced design.

• As the innovative reactor was selected the BN-1200 design.

• Among various reactor types considered, which are currently in operation demonstrating benefits and also specific deficiencies, the BN-600 and BN-800 have been selected as the reference design.

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Design decisions on BN-1200 reactor

• 1 – intermediate heat exchanger (IHX)

• 2, 3 – main and guard vessels respectively

• 4 – support skirt • 5 – pressure chamber • 6 – core catcher • 7 – core • 8 – pressure pipeline • 9 – primary circuit main

circulating pump (MCP-1) • 10 – EHRS heat exchanger • 11 – control rod drive

mechanisms (CRDM) • 12 – rotating plugs

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Main characteristics of BN-1200 power unit

Reactor BN-600 BN-800 BN-1200

Nominal thermal capacity, MW 1470 2100 2800

Gross electric power, MW 600

880 1220

Number of heat generating loops 3 3 4

Primary circuit temperature at IHX inlet/outlet, °C

535/368

547/354

550/ 410

Secondary circuit temperature at SG inlet/outlet, °C 510/318 505/309

527/355

Third circuit parameters: Live steam temperature, °C Live steam pressure, MPa Feed water temperature, °C

505 14 240

490 14 210

510 14 240

Efficiency, gross/net, % 42.5 / 40

41.9 / 38.8 43.5 / 40.7 15

INPRO approach in the area of safety

• INPRO assessment in the area of Safety is

not a Safety Assessment using the IAEA

Safety Standards (as defined in the IAEA

Safety glossary);

• Safety Analyses are necessary prerequisites for

the INPRO assessment in the area of Safety.

• The INPRO assessor needs results of safety

analyses as input to perform judgments

whether INPRO Criteria are met.

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INPRO approach in the area of safety

BP – 4; UR – 14; CR+EP=55.

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Main conceptual ways of reactor safety improvement

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• Minimization of excess reactivity for the fuel

burn-up

• Decrease of sodium void reactivity effect

• Use of passive devices for reactivity control

• Use of passive devices for decay heat removal

• The concept sets as the purpose: exception of reactor runaway on prompt neutrons by limitation of excess reactivity ρburn-up< βeff

Basic principle BP1(Defence in depth)

Basic principle BP1: Installations of an Innovative Nuclear Energy System shall incorporate enhanced defence-in-depth as a part of their fundamental safety approach and ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installations.

The basic principle BP1 contains 7 user requirements UR1-UR7

UR1.1 (robustness): Installations of an INS should be more robust relative to existing designs regarding system and component failures as well as operation.

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Criterion CR1.1.1 (robustness)

Indicator IN1.1.1: Robustness of design (simplicity, margins).

Acceptance limit AL1.1.1: Superior to existing designs in at least some of the aspects discussed in the text.

Evaluation parameters (EP)

EP1.1.1.1: Margins of design

EP1.1.1.2: Simplicity of design

EP1.1.1.3: Quality of manufacture and construction

EP1.1.1.4: Quality of materials

EP1.1.1.5: Redundancy of systems

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EP1.1.1.1 Margins of design Thermal and mechanical fuel design

Construction features Effect produced

1 Enlarged size of fuel pins and fuel assembly

Reduced core power density (up to 1.8 times) Reduced linear heat generation rate Decreasing of neutron flux Increase fuel assembly life-time

2 Gaseous plenum was increased

Accumulation of gaseous fission products

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Usage of reference design solutions for fuel pin and fuel assembly and experimental justification of hydraulic and mechanical performance

Providing of design temperature and mechanical loads

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4 Three types of FA of the same Pu content

Reduce the power peaking factor providing of burn-up flatting , increasing of margin on temperature

5 Stable negative reactivity coefficients, effects and sodium void effect less ef

Improve of the transients Reduce of the frequency of DBA and BDBA

6 Increasing of fuel content Increase of the breading factor reducing of reactivity margin to burn-up, reducing of margin on the control rods and decrease frequency of the reactivity accidents

7 Nitride fuel

EP1.1.1.1 Margins of design Neutronic and thermal hydraulic design of the core

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EP1.1.1.1 Margins of design Neutronic and thermal hydraulic design of the core


8 Increasing of in-vessel ionization chambers number (up to 2 times) Improve the I&C

Reduce of the frequency of reactivity accidents

9 Excluding of neutron guides

10 The CRDM control system prevents simultaneous withdrawal of more than one control rod from the core

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EP1.1.1.1 Margins of design Passive systems


11 Additional passive control rods based on temperature principle of actuation

Improve of the BDBA management

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Emergency heat removal system with air exchanges connected to the primary circuit and with increasing number of trains and check valve in the autonomous heat exchanger

The EHRS insures cool down of the reactor core by sodium circulation arranged directly through fuel subassemblies

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Replacement of bursting disks in the secondary circuit to the passive scheme to flush the water and sodium interaction products under loss of Steam Generation pressure

Elimination of the circuit disconnection under the bursting disk misoperation

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EP1.1.1.1 Margins of design Additional engineering systems and solutions


13 In-core placement of the boron-carbide-based primary shielding

Reduce of the neutron irradiation of in-vessel structures Increase of the reactor service life up to 60 years Decrease of the shielding steel intensity

14 Leak-tight above-reactor premise

Prevent of the radioactive fission products release

15 Core catcher Collection on melted in-core structure

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In – reactor vessel arrangement of systems and pipe-lines with radioactive sodium

Exclude the pipe with radioactive sodium outside of reactor Decrease the frequency of accidents with radioactive release

17 Pipelines safety cover

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Experimental and calculation R&D studies for fuel pin, FA, control rod assembly, temperature passive control rod, CRDM, EHRS, refueling mechanism, neutronic parameters control system, sodium monitoring system, steam generator, main circulation pumps, structure materials

Justification of the design solution Decrease of the frequency of failures for equipment and systems

19 Modernization, development, validation of calculation codes, including uncertainty and sensitivity analysis

Increase of the accuracy of design characteristic calculations

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3D modeling, collective parallel 3D product development technologies , engineering test-simulator, databases of power unit main equipment

Decrease of the frequency of failures owing to Human factor

EP1.1.1.1 Margins of design R&D work to justify the design solutions

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EP1.1.1.1 assessment

• Designing components, systems and structures of reactor facility for external effects of intensity higher than that in the existing analogs.

• Evaluation parameter EP1.1.1.1 has been met by BN-1200 design

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EP1.1.1.2 Simplicity of design


1. Four identical loops

Simplification of loops layout and technological schemes Possibility to use some reference equipment

2.

Elimination of in-vessel shielding owing to boron shield assemblies application

Simplification of in-reactor layout Decreasing of reactor vessel size

3. Elimination of neutron guides

4. Combining of the buffer tank with the MCP-2 tank

Improvement of technical and economic characteristics

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5. Bellows compensators application

Reduction of the secondary sodium pipelines

6.

In-reactor storage to keep the discharged fuel assemblies during 2 years

A simplified BN-1200 reloading system and reduced residual heat of discharged assemblies Elimination of the spent subassemblies (SSA) sodium drum to keep the SSA before their cleaning and transportation to water pond

EP1.1.1.2 Simplicity of design(2)

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7.

Use of vertical once-through shell-and-tube-type steam generators (8 modules) instead of sectional-modular steam generators as in BN-600 (72 modules) and BN-800 (60 modules)

Reduce the material consumption of heat removal loops and overall dimensions of the building

8. Use of the vertical elevator, as well as transfer cell and washing cell combined together

Simplification of the fuel handling system

9. Combination of transfer and washing boxes

Simplification of the system

Evaluation parameter EP1.1.1.2 has been met by BN-1200 design

EP1.1.1.2 Simplicity of design(3)

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EP 1.1.1.3 - Quality of manufacture and construction

• In the design BN 1200 were taken into account the shortcomings that have been in BN-600 and BN-800. The basis of the development project of the BN-1200 consists of the following basic conceptual positions:

The maximum possible use of tested and proven scientific and technical solutions implemented in the BN-600 and BN-800;

The application of new technical solutions to improve the security, power efficiency and fuel efficiency with working and justification of these decisions on the existing and newly established stands;

The same choice of electric power to the NPP-2006 with a view to a common approach to the selection of sites for nuclear power plants and the unification of the generator and other electrical equipment system of issuing of electricity;

Possibility of the transportability of large equipment by rail.


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EP1.1.1.4 Quality of materials

Development of technology, including the improving of chemical content, experimental determination of their performance, justification : - fuel pin materials; - reactor vessel; - reactor internals; - steam generation; - anti-cavitation coatings etc.

Increase of reliability of the equipment and systems

The BN- 1200 used materials that are approved for use in nuclear power.


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EP 1.1.1.5 - Redundancy of systems

• Independent system of emergency heat removal through an air heat exchanger (AHE) added to the basic system of the emergency removal of heat from the reactor through the first and second loop and the steam generator

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Estimation of EP1.1.1.5

• Increased redundancy of systems has been demonstrated in the BN-1200.

• Thus, EP1.1.1.5 has been met. • Final assessment of criterion CR1.1.1 —

robustness of design. The assessment of the evaluation parameters of the first criterion CR1.1.1 has confirmed an increased robustness of the BN-1200 reactor design in several aspects compared with reactors BN-600 and BN-800.

• Therefore CR1.1.1 has been met by the BN-1200 design.

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CR 1.1.2 – Operation • CR 1.1.2 – Operation. • IN 1.1.2 - High quality of operation. • EP 1.1.2.1- Margins of operation. • Rationale. Operator errors during operation BN-1200 will not

lead to severe accidents with core damage and release of radioactive materials due to advanced self-protection properties and BN reactors experience.

• EP 1.1.2.2 - Reliability of control systems • Rationale. Better than BN-600. The reliability of control

systems is based on years of experience of BN-600 and the additional improvement of equipment.

• EP 1.1.2.3- Impact from incorrect human intervention. • Rationale. Operator errors during operation BN-1200 does

not lead to severe accidents with core damage and release of radioactive materials due to advanced self-protection properties.

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BN-1200 REACTOR LAYOUT

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1 – reactor 2 – MCP-2 3 – steam generator 4 – air-cooled HX 5 – main secondary pipelines 6 – EHRS pipelines 7 – bellows compensators

BR-5/10 in 1959

BOR-60 in 1969

BN-350 in 1973

BN-600 in 1980

BN-800 Construction In 2015

BN-1200 design

Experimental reactors

Power reactors

Experience in BN technology development

50 years of successful design and operating experience. Opportunity for

construction of commercial fast reactor BN-1200

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Engineering solutions for BN-1200 safety enhancement

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CR 1.1.2 – Operation(2)

• EP 1.1.2.4 - Quality of documentation.

• Rationale. The quality of the documentation meets with the requirements of regulatory documents.

• EP 1.1.2.5 - Quality of training

• Rationale. Training will be conducted in accordance with the requirements of regulatory documents and experience. Practical experience with BN-350, BN-600 and BN-800 will be used.

• EP 1.1.2.6 - Organization of plant.

• Rationale. Use experience of Russian and foreign nuclear power plants with fast reactors.

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CR 1.1.2 – Operation(3)

EP 1.1.2.7. Availability/capability of plant. Rationale. High capacity factor is ensured by long-term operation BN-600 EP 1.1.2.8. Use of world-wide operating experience. BN-1200 project is based on the experience gained in the global nuclear power industry, as well as resulting in the operation of reactor units BN-350, BN-600 and BN-800. The experience gained will allow for the BN-1200: to exclude the possibility of non-standard situations; to prevent a number of troubles when operating unit; to show the reliability and safety of the blocks BN-1200, the stability and ease of management;

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Final assessment of criterion CR1.1.2 (operation)

The experience of BN-350 reactor has made a valuable contribution to the understanding of the technology of large power plants with fast sodium-cooled reactors and were mostly taken into account in projects of power units with reactors BN-600 and later the bn-800.

The BN-1200 reactor meets requirements of criterion СR 1.1.2.

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CR 1.1.3-Inspection. • CR 1.1.3-Inspection.

• IN 1.1.3-Capability to inspect. • Rationale. The design of the BN-1200 maximum uses

tested and proven scientific and engineering solutions implemented in the BN-600, BN-800, as well as new solutions that improve the ability of inspection of nuclear reactor.

• The inspection program for BN-1200 meets regulatory requirements.

• Inspections for BN-1200 are simplified due to the special measures.

• Estimation of CR 1.1.3. Criterion CR1.1.3 has been met by BN-1200 design.

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CR 1.1.4-Failures and disturbances • IN 1.1.4-Expected frequency of failures and disturbances.

• Rationale. The large number of innovative decisions to reduce the frequency of failures and disturbances were implemented in BN-1200. The following examples shows the most important ones.

• Placing all systems with radioactive sodium within the reactor vessel eliminates the most dangerous class of design-basis accidents with leakage of radioactive sodium.

• Use the safety shell for equipment and pipelines of the second circuit reduces the possibility of a leak of non-radioactive sodium and consequences of such an accident.

• Implementation new technical solutions have allowed to reduce expected number of failures and disturbances.

• Final assessment of criterion CR1.1.4 (failures and disturbances).. Embedded in the project of BN-1200 design solutions that reduce the expected frequency of failures and disturbances in the INES, and the use of experience of operating reactors with sodium coolant allow to make a conclusion about the satisfaction of the project BN-1200 for criterion 1.1.4. 43

User requirement UR1.1 robustness

• Final assessment of user requirement UR1.1 robustness. All criteria were judged to meet their acceptance limit.

• Thus, user requirement UR1.1 is deemed to have been met by the BN-1200 design.

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5.Main problems of the INPRO methodology in area of safety and ways of improving

1. A huge number of criteria!

2. In general the methodology is focused on the problems of the thermal reactor. And this is right! There are many.

3. However, the physics of fast reactors differs from thermal reactors (no poisoning, no hydrogen, no pressure, low reactivity margin, etc.) And not only physics, but also thermal hydraulic!

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Main problems of the INPRO methodology in area of safety and ways of improving(2)

5. Methodology should be focused on the most important points.

6. The number of the criteria should reduce.

7. Description of the criterion and instruction for INPRO assessment must be accurate and simple

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Main problems of the INPRO methodology in area of safety and ways of improving(3)

10.It would be nice to consider an approach based on key indicators

11.The use of methodology in existing form requires a great volume of time, huge amounts of data.

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Conclusion on INPRO methodology

• Improving of methodology moves in the right direction

• The authors have already reduced the number of base principles and user requirements

• But the number of criteria is decreasing small!

• We are trying now to use updated version for BN-1200 reactor assessment

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Conclusion on INPRO methodology

• Features of fast reactor physics is practically not considered in methodology.

• It would be nice to allocate a set of key indicators for the main safety parameters and more simple assessment

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Conclusions

• Russia has the largest experience in the world of research in the field of technology of fast reactors. Three fast neutron reactor- prototype BN-350 and BN-600 and BN-800 were built.

• BN-350 is already stopped, the reactor BN-600 is successfully operated, demonstrating high safety performance and cost-effectiveness, the reactor BN-800 recently was connected to grid.

• The development of the technical design of the reactor BN-1200, is now complete.

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Conclusions

• High technical and economic characteristics, safety characteristics in the BN-1200 design are insured by an optimal combination of reference and new solutions

• Proven scientific and engineering solutions implemented in the BN-600 and BN-800 were used in BN-1200 project .

• Analysis of the reliability and safety of the power unit with BN-1200 reactor was carried out in the framework of the INPRO methodology.

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Conclusions

• The results of the evaluation criteria showed that the BN-1200 meets all the basic principles and requirements of users, incorporated in the INPRO methodology in the "safety" of nuclear reactors. The validity of this assessment is based on a large amount of theoretical and experimental studies, as well as many years of experience operating power fast reactors with sodium coolant.

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Conclusions

• All design, engineering and operational parameters of the BN-1200 were compared with the same characteristics of the existing and designed facilities.

• All criteria were evaluated by the INPRO methodology: 4 basic principles, user requirements 16, 36 criteria and indicators, 21 estimated parameter.

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Thank you very much

for your attention

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