A S T R I D. Safety orientations during ASTRID conceptual design phase IAEA-CN

Similar documents
Safety for the future Sodium cooled Fast Reactors

GENERATION-IV SODIUM-COOLED FAST REACTORS AND THE ASTRID PROJECT

SDC and SDG discussions related to Design Extension Condition [DEC] GIF SDC Task Force Member Yasushi OKANO

NEXT STEP FOR NUCLEAR POWER PLANT GENERATION IV

R&D PROGRAMMES RELATED TO SEVERE ACCIDENT IN SFR

Passive Complementary Safety Devices for ASTRID severe accident prevention

JULES HOROWITZ REACTOR (JHR)

FRENCH R&D PROGRAM ON SFR AND THE ASTRID PROTOTYPE

French R&D program on SFR and the

THE FRENCH FAST REACTOR PROGRAM - INNOVATIONS IN SUPPORT TO HIGHER STANDARDS

Design features of Advanced Sodium Cooled Fast Reactors with Emphasis on Economics

IMPACT OF A PROGRESSIVE DEPLOYMENT OF FAST REACTORS IN A PWR FLEET

Safety design approach for JSFR toward the realization of GEN-IV SFR

KNOWLEDGE PRESERVATION AND DATA COLLECTION ON FAST REACTORS IN FRANCE

COSI6 A Tool for Nuclear Transition Scenarios studies

Gas Cooled Fast Reactors: recent advances and prospects

Nuclear Safety Standards Committee

WENRA APPROACH WITH RESPECT TO DESIGN EXTENSION OF EXISTING REACTORS

Safety Design Requirements and design concepts for SFR

Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor

FROM RESEARCH TO INDUSTRY

Controlled management of a severe accident

SENIOR REGULATORS MEETING Strengthening the Implementation of Defence in Depth IAEA Perspective

ASTRID. Presented by A. VASILE (CEA) ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION

PLATEAU FACILITY IN SUPPORT TO ASTRID AND THE SFR PROGRAM: AN OVERVIEW OF THE FIRST MOCK-UP OF THE ASTRID UPPER PLENUM, MICAS

THE ASTRID PROJECT STATUS, COLLABORATIONS, LESSONS FROM FUKUSHIMA ACCIDENT ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION

IAEA Workshop - Research Reactors. Implementation of the post-fukushima Daiichi accident Enhancement Programme for RRs. Sydney - December 2017

Considerations on the performance and reliability of passive safety systems for nuclear reactors

Applicability of PSA Level 2 in the Design of Nuclear Power Plants

International Review on Safety Requirements. for the Prototype Fast Breeder Reactor Monju

NEW TEXTILE STRUCTURES AND FILM-BOILING DENSIFICATION FOR SIC/SIC COMPONENTS (IV Gen.)

WENRA Approach with respect to Design Extension of Existing Reactors

ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS. Alessandro Alemberti

TECHNOLOGICAL BREAKTHROUGHS AND ELECTRICITY OF TOMORROW

Heterogeneous sodium-cooled fast reactors with low sodium void effect.

The fuel handling route of ASTRID at the beginning of the basic design

REGULATORS FORUM PILOT PROJECT REPORT

MOLY PRODUCTION IN THE JULES HOROWITZ REACTOR CAPACITY AND STATUS OF THE DEVELOPMENT

SAFETY REQUIREMENTS IN FRANCE FOR THE PROTECTION AGAINST EXTREME EARTHQUAKES

FROM PIEZOMETRIC MONITORING TO GROUNDWATER FLOODING RISK MANAGEMENT

Dutch Safety Requirements for Nuclear Reactors: Fundamental Safety Requirements

Passive Safety Features and Severe Accident Scenarios of the small metal-fueled fast reactor system

WENRA views on Defence-in-Depth for new reactors

Stress tests specifications Proposal by the WENRA Task Force 21 April 2011

Post-Irradiation analysis of fission gases in nuclear fuels

Profile SFR-16 DOLMEN FRANCE Saint Paul Lez Durance FRANCE

Implementation of SSR2/1 requirements for Nuclear Power Plant Design in Polish regulation.

WENRA and its expectations on the safety of new NPP

CNSC Fukushima Task Force Nuclear Power Plant Safety Review Criteria

ESSENCE AND CHARACTERISTICS OF THE ATMEA TECHNOLOGY: THE ATMEA1 REACTOR

WORKING MATERIAL. Technical Meeting on Impact of Fukushima event on current and future FR designs

Evolution of Nuclear Energy Systems

Challenges and R&D program for improving inspection of sodium cooled fast reactors and systems

M ertinssafety. The new German Safety Criteria for Nuclear Power Plants in the view of international standards. Prof. Dr. M.

IAEA Workshop. Implementation of lessons learned from the Fukushima Dai-ichi accident Research Reactors. Sydney - December 2017

Basis for the Safety Approach for Design & Assessment of Generation IV Nuclear Systems. Revision 1

IAEA Technical Meeting on Priorities in Modelling and Simulation for Fast Neutron Systems

LFR core design. for prevention & mitigation of severe accidents

Reactivity insertions for the Borax accident in ORPHEE research reactor

POST-FUKUSHIMA STRESS TESTS OF EUROPEAN NUCLEAR POWER PLANTS CONTENTS AND FORMAT OF NATIONAL REPORTS

The Generation IV Gas Cooled Fast Reactor

Severe Accident Countermeasures of SFR (on Monju)

TRANSIENT ANALYSIS OF THE ASTRID DEMONSTRATOR INCLUDING A GAS NITROGEN POWER CONVERSION SYSTEM WITH THE CATHARE2 CODE

ASTRID core Design objectives, design approach, and R&D in support

AFCEN new code for Design of Civil Structures: from ETC-C to RCC-CW

The Nuclear Safety Authority (ASN - Autorité de Sûreté Nucléaire),

Lessons Learned from the Fukushima Dai-ichi Accident and Responses in Regulatory Requirements

Improvements in defense in depth on French NPPs following Fukushima Accidents

Material Selection According to ALARA during Design Stages of EPR. P. Jolivet, A. Tamba, F. Chahma AREVA

ANTICIPATED ANALYSIS OF FLAMANVILLE 3 EPR OPERATING LICENSE - STATUS AND INSIGHTS FROM LEVEL 1 PSA REVIEW

Naturally Safe HTGR in the response to the Fukushima Daiichi NPP accident

Mitja Uršič, Matjaž Leskovar, Renaud Meignen, Stephane Picchi, Julie-Anne Zambaux. Fuel coolant interaction modelling in sodium cooled fast reactors

WASTE MANAGEMENT STRATEGY. French Experience in Legacy Waste Management. 1 st International Forum on the Decommissioning of the Fukushima Daiichi NPS

Arab Journal of Nuclear Science and Applications, 48(3), ( ) 2015

French Nuclear Reactors Reaching 40 Years

PRACTICES FOR NEUTRONIC DESIGN OF RESEARCH REACTORS: SAFETY AND PERFORMANCES

SAFETY ENHANCEMENT TECHNOLOGY DEVELOPMENT WITH COLLABORATIVE INTERNATIONAL ACTIVITY

ITER at Cadarache : An Example of Licensing a Fusion Facility

EDF NPPs Post-Fukushima Complementary Safety Assessments

Extension to 60 years lifetime

Safety Classification of Structures, Systems and Components in Nuclear Power Plants

Highlights From the Work of the NEA on Impacts of the Fukushima Accident. Javier Reig Head, Nuclear Safety Division

Improvements on French Nuclear Power Plants Taking Into Account the Fukushima Accidents

Safety Requirements for HTR Process Heat Applications

WM2014 Conference, March 2 6, 2014, Phoenix, Arizona, USA

Safety Provisions for the KLT-40S Reactor Plant

From the Accident at the Fukushima Daiichi NPS: Efforts to Improve Safety

Assessing and Managing Severe Accidents in Nuclear Power Plant

French Strategy For GENIV reactors

Safety Challenges for New Nuclear Power Plants

Harmonized EUR revision E requirements corresponding to currently available technical solutions

International Atomic Energy Agency. Impact of Extreme Events on Nuclear Facilities following Fukushima. Dr C H Shepherd Nuclear Safety Consultant, UK

Background. Introduction. Overview of vendor design review process

Gas Cooled Fast Reactor

State of the Art and Challenges in Level-2 Probabilistic Safety Assessment for New and Channel Type Reactors in India Abstract

TS1-5: Revision of NS-G-2.15 and Its Implementation for Verification and Validation of Severe Accident Management Guidelines

Attachment VIII-1. July 21, 2011 Nuclear and Industrial Safety Agency

Modelling of the contamination transfer in nuclear reactors: The OSCAR code Applications to SFR and ITER

Joint ICTP-IAEA Essential Knowledge Workshop on Deterministic Safety Analysis and Engineering Aspects Important to Safety. Trieste,12-23 October 2015

NUCLEAR FUEL AND REACTOR

Transcription:

A S T R I D Advanced Sodium Technological Reactor for Industrial Demonstration International Conference on Fast Reactors & Related Fuel Cycle March 4-7, 2013, Paris, France Safety orientations during ASTRID conceptual design phase P. LO PINTO, R. DOUSSON, J.C. ROBIN (CEA) B. CARLUEC, S. EHSTER-VIGNOUD, S. BEILS (AREVA) P. MARITEAU, F. GIFFON (EDF) IAEA-CN-199-267 www.cea.fr DEN/CAD/DER/CPA Pierre Lo Pinto astrid@cea.fr 1ER MARS 2013 CEA 10 AVRIL 2012 PAGE 1

CONTENTS 1. Introduction Presentation of the Safety Orientations Document (DOrS) Global safety objectives 2. Basic components of the safety orientations Specific risk diagram of ASTRID Implementation of design safety methods Lines of Defense and Lines of Mitigation methods 3. Implementation of safety orientations through the conceptual design To promote natural behavior of the plant New approach of core severe accident Decoupling between CDA study and lines of Mitigation design 4. Other notions contributing to robust safety demonstrations Safety demonstration for practically eliminated situations Progressiveness of the approach Example of progressiveness : Subassembly fault family 5. Concluding remarks on ASTRID safety orientations CEA FR13 06 March 2013 PAGE 2

1.1 INTRODUCTION The current phase of ASTRID project is devoted to the choice of the most structuring options for the conceptual design In order to integer earlier the safety concerns into the design project, a Safety Orientations Document (DOrS) was delivered in 2012 with a double purpose : To define the need of assessment studies for selecting the design options from safety viewpoint, To initiate the exchanges with the licensing authority before the selection of the most structuring design options This presentation gives some information on major safety orientations specific to the ASTRID project CEA FR13 06 March 2013 PAGE 3

1.2 GLOBAL SAFETY OBJECTIVES Consequences levels and probabilistic targets : < 10-5 /year Prevention of Severe Accident with core meltdown, including whole events and hazards Radiological releases not requiring off-site countermeasures < 10-6 /year Mitigation of Severe Accident Unacceptable radiological releases but consistent with off-site countermeasures i.e. postponed, limited in time and area < 10-7 /year Prevention of massive or early radiological releases i.e. not consistent with efficient off-site countermeasures Approach basically deterministic for a better safety implementation through the conceptual design 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 4

2.1 RISK DIAGRAM OF ASTRID A specific approach beyond the design basis domain : Definition of SP, SM and SPE domains and related analyses rules Classification not based on frequency range but on level of degraded plant state With the objective to class in SP prior to considering SM and at least SPE 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 5

2.2 IMPLEMENTATION OF SAFETY DESIGN METHODS To translate the safety principles into practical analyses tools a good way to get robust safety demonstrations In addition to the existing French regulatory fundamental safety rules, some examples: Method of «Lines of Defense» (from SFR feedback) New concept of «Line of Mitigation» method Type of demonstration for practical elimination of some situations (SPE) New definition of safety classes for important equipment (SSC) Appropriate methods for specific SFR events (ex. LBB implementation) Definition of «hard core» provisions (Fukushima feedback) 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 6

2.3 «LOD» & «LOM» METHODS S.A. Prevention S.A. Mitigation Method : Lines of defense (LoD) Lines of Mitigation (LoM) Approach type : «Bottom-Up» «Top-Down» Objective : Probabilistic targets Consequences reduction Lines validation criteria : Demonstration : Application domain : Safety classification of SSC : Number of lines, reliable, independent, common mode absence Equivalent to 2 strong + 1 medium lines Prevention including SPE Complementary to analysis by function Equipment ensuring all functions of one LoM. Each LoM homogeneous: approach weak link of chain Minimization of radiological release with decoupling approach Complementary to analysis by barrier method Complementary to analysis by function «Hard Core» contents: One LoD per SPE All equipment involved in one same LoM 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 7

3.1 TO PROMOTE NATURAL BEHAVIOR OF THE PLANT Objective is not to substitute natural behavior for safety systems but to improve the safety level by additional diverse safety provisions : Enhanced natural behavior (i.e. unprotected transients) as a backup of the safety systems To complete the part brought by the natural behavior by complementary safety devices if needed (ex. CSD for achieving a final safe state) To promote favorable natural behavior both : in SA prevention domain, as a third defense level in SA mitigation domain, in order to reduce the potential consequences and then to less attack the safety mitigating provisions Improvement of the natural behavior concerns all safety functions (reactivity mastery, DHR, confinement ) against all type of initiating events families 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 8

3.2 NEW APPROACH of CORE SEVERE ACCIDENT CDA studies from different events families (initiating transients) with identification of : typical core degraded states shared by different scenarios (crosscut states) key parameters leading to a range of consequence results. Operating conditions Initial states Initiating transient ULOF UTOP USAF Primary phase core geometry loss TOP from lower power, high T TOP from over power TOP from nominal state Transition phase Dispersioncompaction Delayed compaction Delayed dispersion compaction Secondary phase Criticality Corium relocation final safe state Dynamical ejection Thermal ejection Some features of the new approach : Taking account of SA despite high reliability of safety systems and natural behavior contribution in prevention CDA studies not based on only one scenario but from different events families Objective of non energetic CDA by conceptual core design and CSD if needed Decoupling between CDA results and lines of mitigation design 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 9

3.3 DECOUPLING between CDA STUDY and LOM DESIGN Approach adapted to get robust mitigation countermeasures : «Top-Down» approach through the «Lines of Mitigation» method Implementation of the Defense-in-Depth level 4 (mitigation provisions) should prevent a common mode fault into the approach; for this purpose it is recommended : As regards the containment : the reactor should be designed so that any scenario of core degradation cannot lead to a high mechanical energy release. Nevertheless, components and structures required to mitigate CDA consequences, should be designed to withstand, as far as reasonably feasible, against a hypothetical mechanical energy release. As regards the confinement : even if the source term mobilized by a SA scenario involving core meltdown might be limited, the design provisions related to the confinement function should be optimized as far as reasonably feasible. 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 10

4.1 SAFETY DEMONSTRATION FOR SPE To use a safety demonstration method suitable for practical elimination of some situations (SPE) Situations that could lead to a massive or quick radiological off-site release (i.e. not manageable by countermeasures) SPE stemming from possible «cliff edge» effect on consequences or from SA scenario without possible efficient mitigation provisions SA Prevention LoD : 2a + b 3 LoD SA Situations SM Beyond some level of consequences SPE Provisions devoted to consequences mitigation (catcher, confinement ) Evidence of sufficient provisions devoted to reduce the situation frequency (deterministic and 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 11 probabilistic ) Deterministic approach completed by probabilistic insight : at least equivalent to 3 lines of defense with common mode resistance and high confidence level

«DB» 4.2 PROGRESSIVENESS OF THE APPROACH Progressive escalation by events family Slow or fast LOF Slow or fast TOP SAF «SP» DB initiators with additional failures including safety systems (Uxxx) Initiators more severe than DB initiators (ex. postulated fuel assembly melting) «SM» S.A. scenario from SP by additional aggravating hypothesis per family Generic approach not connected to a reference scenario to be justified «SPE» Safety demonstration based on robust prevention provisions (3 LoD) As for situations in continuation of SP, the mitigation measures (SM) could have favorable effects RR High energetic Severe Accident Massive or early radiological off-site release 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 12

4.3 EXAMPLE OF PROGRESSIVENESS: SAF FAMILY SAF family : a new strategy as regards the postulated «fuel assembly meltdown» Previous SFR approach (TIB) Total & Instantaneous Blockage scenario with detection and protection No other case of local fuel melting considered except for an unprotected control rod withdrawal (CRW) New approach for ASTRID Progressiveness considering various events from a «partial fuel assembly blockage» without melting towards the «global core meltdown» situation : Exhaustive sensitivity study on efficient detection-protection means Knowledge and understanding of physical evolution of different cases of fuel assembly blockage (size and delay) Tacking account of global core meltdown (SM) from the SAF family with the same joint objective : no energetic CDA 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 13

5. CONCLUDING REMARKS ON SAFETY ORIENTATIONS In comparison with previous SFR, safety improvement is expected through the conceptual design by implementation of ASTRID safety orientations. Some of them are : Appropriate treatment of local faults (detection, progressiveness ) Approach by events family for both prevention and mitigation of SA Enhanced inherent plant behavior as a third prevention level of SA Generic approach of CDA considering : all types of initiating transients, typical degraded core states, key parameters leading to a range of results New concept of lines of mitigation method (LoM) Decoupling between CDA results and design of SA mitigation provisions facing : Hypothetical mechanical energy release, Potential radiological source term. Rational demonstration of practically eliminated situations (SPE) Integration of Fukushima lessons through hazards concerns beyond the Design Basis, including the hard core notion (see dedicated presentation during FR13). 1ER MARS 2013 CEA FR13 06 March 2013 PAGE 14

PAGE 15 CEA 10 AVRIL 2012 Commissariat à l énergie atomique et aux énergies alternatives Centre de Cadarache 13108 Saint Paul lez Durance Cedex T. +33 (0)4 42 25 32 54 F. +33 (0)4 42 25 20 13 DEN DER CPA (astrid@cea.fr) 1ER MARS 2013 Etablissement public à caractère industriel et commercial RCS Paris B 775 685 019