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

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1 Safety design approach for JSFR toward the realization of GEN-IV SFR Advanced Fast Reactor Cycle System R&D Center Japan Atomic Energy Agency (JAEA) Shigenobu KUBO

2 Contents 1. Introduction 2. Safety design approach of JSFR 3. Safety design concept of JSFR Design against ATWS type events Design against LOHRS type events Containment 4. Conclusion ATWS: Anticipated Transient Without Scram, LOHRS: Loss Off Heat Removal System 2

3 Introduction Design targets for Generation IV reactors; Excel in safety and reliability, sustainable and competitive energy generation, proliferation resistance and physical protection In order to achieve cost competitiveness as base load energy source, commercial SFRs shall cover large output power at least equivalent to current LWRs. For the last decades SFRs have been designed, constructed and operated worldwide to obtain important experiences related to reliable operation and responses to accident conditions in the involved countries. Safety research and development works have been done and on going in the related field such as core fuel performance, thermal hydraulics, molten fuel and radioactive materials behavior and sodium chemical reaction. Safety design of GEN IV SFR can be established on these accumulated technical basis. 3

4 Introduction In the design and evaluation of SFRs including Monju core disruptive accidents (CDAs), which are typical severe accidents of SFR, has already been addressed. And it is important to assure the measures against severe accidents reflecting the lesson learned from the TEPCO s Fukushima Dai ichi NPS accident. The safety design criteria for Generation IV SFR [GIF SFR SDC] is developed taking these aspects into account and important basis for JSFR safety design. 4

5 Basic idea for securing safety Basic safety principle is Defence in depth (DID) which is same as LWRs. In addition to conventional three defence lines so called prevention of abnormal occurrence, prevention of progression of abnormal conditions and control of accidents within design basis, control of severe plant conditions shall be required. GEN IV SFR aims at Elimination of the need for offsite emergency response by enhancing prevention and mitigation of significant core damage in the severe plant conditions. DID Level 1 Level 2 Level 3 Level 4 Reinforced! Conventional parts Prevention & Mitigation Design Extension Conditions (DEC) 5

6 Safety design approach of JSFR Comprehensive safety approach = Active engineered safety systems + Natural behavior (inherent or passive feature) to terminate SA + Accident Management Proven technology based on the experiences of existing SFRs including Monju is fundamental. Passive reactor shutdown and core cooling capabilities are incorporated as built in manner to prevent core damage in the severe plant conditions. Mitigation of core damage situations within reactor vessel (IVR: In Vessel Retention) is provided, which is based on inherent material relocation and cooling. Accident management measures to be established for Monju will be adequately introduced. These design and operation measures are provided for Practical Elimination of accident situations that could lead to a significant and sudden radioactive release due to a possible cliff edge effect. 6

7 Safety design approach of JSFR Comprehensive safety approach = Active engineered safety systems + Natural behavior (inherent or passive feature) to terminate SA + Accident Management DID Level 1 Level 2 Level 3 Level 4 Reinforced (1) Prevention of abnormal operation & failures Rational design margin Quality assurance Preventive maintenance (Inspection, On line monitoring, and so on) (2) Control of abnormal operation & detection of failures Reactor Shutdown Decay Heat Removal (3) Control of accidents within design basis AOO/DBA Active shutdown systems Prevention Multiple decay heat removal systems with natural circulation Accident management measures (4) Control of severe plant conditions Design Extension Conditions(DEC) Passive shutdown Mitigation Containment IVR Prevention of severe re-criticality Stable cooling of degraded core Inherent mechanism Alternative Cooling 7

8 Difference between ATWS and LOHRS Since the event progressions of ATWS and LOHRS are very different, design measures shall be determined according to their characteristics. ATWS Reference Power or Temperature 1.0 High power core melting Power of Temp. Temp. LOHRS ~2700 Fuel Melting ~1400 Cladding Tube Melting 900~1000 Coolant Boiling 700~800 Creep Rupture of Reactor Coolant Boundary Power Low power core melting after core uncovering ~minutes ~hours ~10hours Time ATWS: Anticipated Transient Without Scram, LOHRS: Loss Off Heat Removal System 8

9 Design against ATWS type events AOO & DBA Inherent reactivity feedback characteristics with negative power coefficient. Operation temperature range: sufficiently below the coolant boiling temperature The safe reactor shutdown: 2 active reactor shutdown systems DEC Passive shutdown capability : SASS (Self Actuated Shutdown System) Mitigation of core damage: Prevention of severe mechanical energy release: limitation of core sodium void worth and molten fuel discharge capability by FAIDUS (Fuel Assembly with Inner Duct Structure) In Vessel Retention: in vessel core catcher 9

10 Self Actuated Shutdown System (SASS) Control rod drive line Fuel assembly Control rod Iron core Curie point type passive CR insertion mechanism Temperature sensitive alloy (Ni-Co-Fe) Effective for all types of ATWS, i.e., ULOF, UTOP and ULOHS Magnetic path One dimensional movement in a subassembly scale reduces the uncertainty for the behaviors during ATWS Coil The de-touch mechanism is used both for active and passive operations and thus Detach surface the safety function can be easily demonstrated and verified during the operation Robust core restraint structure ensures control rod insertion. (rod jamming can be eliminated from cause of reactor shutdown system s failure) The amount of the negative reactivity insertion is large enough to shut core down. 10

11 Improvement of Control Rod Insertion Deformation condition Normal condition Concept of core restraint Fuel subassembly Reactor vessel Core barrel Core support structure Reactor core Normal condition SASS Offset condition Reactor core and its upper structure Flexible Joint Control rod Since core restraint structure geometrically restricts core deformation by heat, irradiation, and earthquake, insertion of control rod is ensured. Even if the core becomes deformed a little, the flexible joint structure protects mechanical jamming of a control rod. Reactor shutdown is ensured even in the severe earthquake. 11

12 Mitigation of core damage IVR concept for ULOF: Unprotected Loss of Flow Phases Initiating phase Early-discharge phase Material-relocation phase Heat-removal phase Required Condition s Design Measures No severe energetics caused by coolant voiding Suppression of maximum void worth less than 6 $ No severe energetics caused by large-scale fuel compaction in molten-core pool Installation of FAIDUS for the molten-fuel discharge before the formation of molten-core pool No severe energetics caused by the motion of core-remaining materials, & no thermal failure of RV caused by the contact of discharged molten materials Enhancement of fuel discharge through primary CRGT, & design optimization of inlet/lower plenums for quenching discharged molten materials No failure of stable cooling avoiding the excess of coolable-limitation thickness in debris bed Installation of multi-layer debris tray to enhance the debris dispersion and to suppress the debris thickness below coolable limitation 12

13 Design against LOHRS type events Since various measures can be used during the longer time period by the core melt, the design measures shall be such that significant core damage is practically eliminated. Natural circulation DHR for the decay heat transport to the atmospheric air Hardly affected by tsunami and flooding Available in case of loss of AC power Reinforcement of air stacks for strong wind, physical barriers and separation for external missiles Accident management: manual adjusting the damper opening to maintain the cooling performance even under DC power depletion for instrumentation and control Alternative cooling measures: gas cooling of the water side of steam generators, additional cooling circuit to the sea water 13

14 Design against LOHRS type events As long as heat is generated from a reactor core, the heat can be removed by the natural circulation of coolant even under loss of electric power supply. Basic performance of natural circulation cooling has been demonstrated by JOYO etc. Available in case of long-term loss of AC power, severe tsunami and flooding DRACS PRACS PRACS: Primary Reactor Auxiliary Heat sink(air) Cooling System DRACS: Direct Reactor Auxiliary Cooling System Ensuring natural circulation force by difference in height Heat generation (Core) 14

15 Design against LOHRS type events Example of alternative cooling Additional cooling circuit to sea water DRACS Physical separation & protection barrier PRACS PRACS: Primary Reactor Auxiliary Cooling System DRACS: Direct Reactor Auxiliary Cooling System Gas Manual adjustment of damper opening Gas cooling of SGs Sea Water Steam Generator Alternative cooling EsL 1 EsL 2 Reactor Vessel Lower coolant intake [EsL2] Cooling case in of siphon case break of loop break Primary Sodium Pump/IHX Secondary Sodium Pump Gas Blowers Alternative cooling Sea Water 15

16 Containment Design measures to prevent the following situations are considered. large scale sodium leak and fire deflagration/detonation of accumulated hydrogen induced by sodium concrete reaction or fuel debris concrete interaction mechanical energy release as a consequence of recriticality Prevention of mechanical energy release and fuel debris concrete interaction : prevention of severe recriticality and IVR Prevent of large scale sodium leak and fire and deflagration/detonation of accumulated hydrogen: double boundary concept, Prevention of vessel melt through against LOHRS type events Load condition: decay heat from the gaseous and volatile fission products dispersed in the containment vessel, and latent heat from leaked sodium. 16

17 Containment AOO & DBA Introducing double boundary design, etc., sodium leak and fire in the reactor building is prevented in DBA category. Therefore, compact and simple SCCV structure can be adopted to the JSFR containment design. DEC Introducing following design, severe mechanical and thermal loads inside CV due to severe accident phenomena are prevented in DEC category. ATWS type DEC : Introduction of re critically free. LOHRS type DEC: Core damage and subsequent core melt through is practically eliminated. Considering that the CV is final barrier, design measures against following challenge factors have been studied. Sodium leak and fire by double boundary failure Heating of CV atmosphere by gaseous and volatile FP, etc. 17

18 Containment The JSFR reactor building is steel plate reinforced concrete structure, and the section designated for primary coolant system is designed as containment vessel. Steel Plate Containment Vessel Concrete Steel Plate Steel-plate reinforced concrete structure 18

19 Conclusion Safety Design Approach for JSFR Comprehensive approach based on the technology matured through worldwide experiences of SFRs and incorporates natural behavior to terminate severe accidents Based on the safety design criteria for Generation IV SFR Safety Design Concept of JSFR For failure to shutdown: Passive shutdown capability, Mitigation of core damage (Prevention of severe mechanical energy release, In Vessel Retention) For failure to remove heat: Prevention of significant core damage (Natural circulation DHR, Alternative cooling measures) Containment: Prevention of severe loads by design measures (IVR, double boundary concept, Prevention of vessel melt through against LOHRS type events) 19

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