GENERATION IV NUCLEAR REACTORS Preliminary safety considerations on SFR GEN-IV Prototype G.B. Bruna...
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Transcript of GENERATION IV NUCLEAR REACTORS Preliminary safety considerations on SFR GEN-IV Prototype G.B. Bruna...
GENERATION IV NUCLEAR GENERATION IV NUCLEAR REACTORSREACTORSPreliminary safety considerations on SFR GEN-IV PrototypeG.B. Bruna IRSN/DSR
VHTR
TRISO
SFR
SUMMARY
I. Introduction
I. GIF Framework
II. Objectives for the GEN-IV Systems
III. Insight on the French strategy
IV. Expectations for the safety demonstration
II. Overview of the main features of the GEN-IV reactor concepts focusing on :
I. A short description
II. Advantages & Drawbacks
Focus on the SFR concept and its safety features
Reference : document on the web site of the IRSN :
« GENERATION-FOUR (GEN-IV) REACTORS / SUMMARY REPORT / MARCH 2007 »
[www.irsn.org/en/document]
Others public references:
www.gedeon.prd.fr / GEDEPEON May 2007
www.physor2004.anl.gov/PlenarySessions.htm / PHYSOR, Chicago, 2004
Parameters to account for …
Energy marketFinancial risk
Environmentprotection
Public acceptance
The designer
?
GIF Framework
Objectives and Context for the GEN-IV Systems
Hazards
Objectives and Context for the GEN-IV Systems
Design and operating feedback
Safety issues derived from licensing process
New safety objectives, new standards…
Specific issues for GEN-IV new systems
(technological orientations,
challenges, etc.)
Should they exist !
It is difficult to go beyond the main safety principles
Requirements: « what it is wished and what it ispossible: that’s the question » !
Participants in the G.I.F.
Russian ConfederationPeople’s Republic of China
Objectives for the GEN-IV Systems
Economic competitiveness
Competitiveness of the nuclear KWh cost, vs fossil energies
Sustainability Increased reactor lifetime (over 60 years)
Optimization of fissile material inventory
Decrease of the waste volume and storage costs
Safety Very low probability of severe damage of the core
No need for off-site emergency plan for severe accidents
Resistance to proliferation and to acts of malicious damage
Fuel cycle minimizing the production of weapon-grade materials
Efficient protection against internal and external hazards
ENHANCED SAFETY
Objectives for the GEN-IV Systems
Reduction of fault rate of normal operation equipments,
Increased protection against external attacks and hazards (plane crash, malevolence, etc.),
Very low probability of major core damage:
Design features
Passive protection system,
No need for off-site emergency plan for severe accidents
According to a defense in depth design approach including severe accidents in the design basis.
SUSTAINABILITY
Optimization of the uranium resources and the fissile material inventory (closed fuel cycle),
Decrease of the waste volume and storage costs,
Waste management taken into account in the design,
Multi-functionality (hydrogen, electricity, industrial heat).
Objectives for the GEN-IV Systems
Generation IV Systems
HTR/VHTR
SFR
GFR
LFR
MSR
SCWR
MSRLFR
SFR GFR
SCWRHTR/VHTR
Insight on the French Strategy
January 2006, impulse of President J. Chirac for the operation of a GEN-IV reactor prototype by 2020
June 2006, adoption of a new Law on the management of radioactive materials and waste, with two milestones:
2012: definition of an industrial scenario for GEN-IV and ADS system
2020: operation of a GEN-IV prototype
Insight on the French Strategy
December 2006, decisions of the French Council of Ministers, and of the « Atomic Energy Committee » including different representatives of the French Government (Research, Industry, Environment, etc.):
involvement of France in the design of GEN-IV systems, in the aim of an industrial deployment in the ‘40
priority to the fast reactor systems allowing a closed fuel cycle
support to the industry for advanced VHTR system design
Insight on the French Strategy
Current Systems Evolutionary Advanced and Revolutionary
About Licensing in France
Planning proposed by CEA for the GEN-IV prototype:
SFR : 2010 : « principles of innovative safety options »
2012 : « proposal of a set of options for the prototype »
GFR : 2009 : « evaluation of safety options »
2012 : « safety report »
Safety Authorities
Operator
5 Authorizations
3
Technica
l
rela
tions
4Technicalassessments
Generalists: power reactors fuel cycle facilities experimental reactors waste
Specialists: mechanical engineering hydraulics, thermal
eng. reactor control I&C severe accidents human factors neutronics seismic studies, etc.
Request forms
2
Integrating knowledge, summaries Research and development requirements
Applications
1
A part of the IRSN’s assignment is to serve as the TSO
for the French Nuclear Safety Authority
Licensing procedures
DISMANTLING
FINAL SHUTDOWN
FINAL OPERATIONOPERATINGCONSTRUCTIONDESIGN
OPERATING LICENCESDISMANTLING
DECREE
FSARPSARPrSA
RSOR FSAR
GOR
EP
GSSR
EP
SOR: Safety Option Report PSAR: Preliminary Safety Analysis Report
GOR: General Operating Rules PrSAR: Provisional Safety Analysis Report
EP: Emergency Plan FSAR: Final Safety Analysis Report
GOR
Authorisation DECREE
EP
General expectations for the safety demonstration of GEN-IV systems
Enhanced safety compared to GEN-III and GEN-III+ (EPR, AP1000, etc.) At least equivalent criteria (probabilistic approach for severe core damage) and confidence level in the safety demonstration
From the IRSN point of view, the current safety approach must be adopted:
defense in depth principle,
« deterministic » approach, supported by extended PSA insight (including “safety margins” assessment)
For systems which have been already built and operated (such as HTR, SFR, LFR, etc) design and operating experience, must be accounted for by the designers to increase the safety.
« Safety margins » Approach
A’
AB
B’
C
Loss of Safety margins
Risk
Definition of the “Risk Space” and its sensitivity to NPP design changes
General expectations for the safety demonstration of GEN-IV systems
General expectations for the safety demonstration of GEN-IV systems
The demonstration of the exclusion of events consequences of which are not accounted for in the design (« practically eliminated »):
big graphite fire in VHTR, big sodium fire in SFR ?
complete break of pipes in VHTR, SFR, GFR ?
melting of the core for TRISO type fuel, for SFR, GFR ?
Which kind of demonstration (« lines of defence », PSA, …) and which confidence level ?
General expectations for the safety demonstration of GEN-IV systems : main challenges
The definition of the most severe accident retained in the BDBA scope, with dedicated safety systems, the prevention, the mitigation of the consequences, mainly concerning the containment/confinement
Co-generation: the safety approach retained for coupled VHTR and industrial installations for production of industrial heat, hydrogen, etc.
What are the events generated by industrial facilities which must be taken into account as operating conditions or external hazards in the safety assessment ?
Does it exist a safety assessment for such facilities?
Is their safety assessment consistent with the assessment for nuclear reactors ?
General expectations for the safety demonstration of GEN-IV systems
Ambitious targets for the radioprotection and the radiological consequences for the public and workers in operation (DBA, BDBA) and in presence of hazards,
A clear identification of
the containment/confinement barriers and safety systems,
their functional requirements vs. operating, incidental and accidental conditions and hazards
General expectations for the safety demonstration of GEN-IV systems : main challenges
The core characteristics, mainly those involved in thesafety studies (neutronics, feedback coefficients, etc.)
The study of the most severe reactivity accidents (if not included in the BDBA): prevention, detection andconsequences
The extensive use of PSA, with topics which need improvements (data bases for equipment, reliability for passive systems, probability evaluation for rare hazards, etc.)
PART II
Main features of the reactor concepts
Maturity of concepts
Advantages & Drawbacks
Summary Review (1/2)
Concept Safety /Safe natural behaviour
Uranium resource
using
Quantity / Waste management
SFR No + + (if multi-recycling)
GFR No (« semi-passive » for
some accidents)
+ + (if multi-recycling)
HTR/VHTR Yes - or = - (graphite, production of plutonium and M.A.)
SCWR No ? ?
LFR Yes (for medium powers)
+ + (if multi-recycling)
MSR Yes (severe accidents to
explore)
Using of Th, more abundant
than U
++ (but risks dues to the salts traitment)
Elements of comparison between GEN-IV concepts
Summary Review 2/2
No system is able satisfying all the GIF criteria
The six concepts do not enjoy the same maturity level : the SFR and the HTR enjoy the most advanced technologies
The VHTR does not permit a closed fuel cycle (as far as current designs and technology are concerned), it needs enriched uranium fueling, but shows some major advantages:
a resistant barrier around the fuel, a safety founded on a natural behavior of the reactor, a capacity to be coupled to industrial processes (heat, H2, etc.)
The SFR allows a closed fuel cycle and enjoys: a proved technology, a widespread operating experience, nevertheless it needs some major improvements (neutronics, risks
dues to the sodium, ISI, etc.)
SCWR
Westinghouse concept (INEL)
Power: thermal / electric
3575 MW / 1600 MW
Temperature of the water: inlet / outlet of the core
280°C / 500°C
Fuel U; enrichment: 5%
Pressure of the water
250 bar
BU 45 GWj/t
Westinghouse concept (INEL)
ADVANTAGES:
Direct conversion cycle: the vapor which enters the turbine is produced into the core (no benefit for the safety)
Fast neutrons (breeder reactor) or thermal neutrons concept
INCONVENIENTS and/or INUMBENT DIFFICULTIES:
The heat exchange between the fuel and the water is not uniform Great uncertainties on the fuel cooling (especially for the super-critical water)
Difficulties with the core design and layout: need for multi-enrichment zones
At the stage of feasibility studies / Non nuclear design and operating feedbacks
MSR
Molten Salt Reactor
Power: 1200 MWth
Coolant: fluor salts, without significant pressure
Moderator: graphite
Fuel: thorium dissolved in the salts, also with U233
Coolant temperature at the outlet of the core: 850°C
Molten Salt Reactor
ADVANTAGES:
No risk of core melting !
Possible on-line extraction of the PFs low consequences in the case of salts leakages
Thorium fuel: abundant « fertile » material
Less waste produced by MWe
Molten Salt Reactor
INCONVENIENTS and/ or INCUMBENT DIFFICULTIES:
Salts are corrosive ( non-metallic materials) and the solubility of the PFs is various in the salts
Melting temperature of the salts > 500°C
Irradiation of the primary circuit structures
Neutronics: complexity (fissions in all the primary circuit !)
This concept has not passed the experimental stage
LFR
Lead Fast Reactor
Power: 25 to 1200 MWth
Coolant: molten lead (or lead-bismuth)
Moderator: none
Fuel: U238+Pu (nitride or metallic type)
Coolant temperature at the outlet of the core: 550°C to 800°C
Alternatives:- Pile without reloading
-Integrated power reactor
-Loop power reactor
ADVANTAGES:
The reactor can be operated with natural or depleted uranium
Lead boiling is almost impossible (T>2000°C)
No strong pressurization
Passive behavior in case of accidental transients reactor control without immediate acting of protective systems or operators
Good compatibility with water (secondary coolant), and no fire risk with air
INCONVENIENTS and/or INCUMBENT DIFFICULTIES:
Molten lead is very corrosive (pumps, clad, vessels, etc.)
Difficulty to wash and decontaminate the equipment immerged in the lead maintenance ?
Significant hydrodynamic pressures (BREST: ~ 1,6 bar)
ISI: not possible for internal structures (BREST)
Difficulties for an core unloading, in case of emergency ?
Activation of lead and bismuth: production of long life waste
Not satisfying operating feedback (submarines)
Maturity: not advanced
HTR/VHTR
High or Very High Temperature Reactor
Power: 300 à 600 MWth
Coolant: Helium, under pressure (dozens of bars)
Moderator: Graphite
Fuel: Uranium low enriched (8 to 15%); pebbles or compacts
Coolant temperature at the outlet of the core: 850°C to 1000°C or more
HTRs : Design and Operating Experience
Peach Bottom - US (1966-1973) :
U/Th / 40MWe / 750°C (T° He)
Fort Saint Vrain - US (1976 – 1989)
840 MWth / 750°C (T° He)
AVR – Jülich – RFA (1966 – 1987)
46 MWth / 850°C (T° He)
THTR 300 – RFA (1985 – 1989)
750 MWth
US Experience
Validation of the concept of coated particles (two layers)
Qualification of the systems for the primary helium purification
Testing of materials under the flux
Peach Bottom 1 – 1966 -1973
U/Th (high enrichment)
40 MWe
Helium : 350°C / 750°C
Good behavior of the fuel
Neutronics disturbances of the core dues to helium bypass (lateral movements of graphite blocks)
Water ingress (failure of the fans bearings) graphite damaging
Anticipated final shut down
Fort Saint-Vrain – 1976-1989
842 MWth et 330 MWth
Spherical coated particles and put into hexagonal graphite blocks
Helium : 350°C / 750°C
US Experience
German Experience
120 000 hrs of operation with a high availability factor (66,4%)
Small doses for the workers during the maintenance
Tests of non protected transients
Fuel not reprocessed AVR (KFA – Jülich) / 1966 -1987
P = 46 MWth / 15 MWe
Tmax helium : 850°C (up to 950°C in 1974)
AVR
HTR-10
HTRs : Design and Operating Experience
HTTR
China / 10 MWth / pebblesTsinghua University
Japan / 30 MWth / prismatic blocksOarai
New low power experimental reactors
Technical Orientations for New Power Reactors Plants Technical and Technological Challenges
EUROPE: ANTARES (VHTR- 600 MWth), + R&D project RAPHAEL
SOUTH AFRICA: PBMR (400 MWth pebbles)
RUSSIAN FEDERATION: GT- MHR (600 MWth prismatic blocks of compacts)
JAPAN: GTHTR-300 (600 MWth prismatic blocks of compacts)
CHINA (Chinergy): HTR-PM (195 MWe)
The VHTRThe VHTR
The VHTR is seen as more efficient than reactors in operation in several aspects:
- A higher thermodynamic efficiency and a wider scope of applications, because of the very high temperature gas supply,
- A different commercial approach, to serve the market segment of medium-scale electricity production, as opposed to the traditional nuclear plants for large-scale production of electricity.
- A minimized environmental impact owing to the robustness of the fuel that retains fission products under both normal and accidental conditions,
- A better resource utilisation and a contribution to waste minimization owing to
Its thermal efficiency, Its quite high burn-up, Its large capacity to transmute Actinides [both Plutonium and Minor
Actinides].
Core design
“ The VHTR Core: A Very Heterogeneous System”
Core heterogeneities (1) [www.physor2004.anl.gov]
Core heterogeneities (2) [www.physor2004.anl.gov]
ANTARES ( free references : [www.areva-np.com] and [www.iaea.org]) [www.physor2004.anl.gov]
Primary circuit and exchangers
Reactor vessel
Helium fan
Isolating valves
Wall of the reactor building
Plates type exchanger
Heat removal system after reactor shut down
Control rod penetrations
ADVANTAGES:
Resistant first barrier up to 1600°C
Very low power density (few MW/m3)
Large inertia due to the important quantities of graphite; fuel temperature 1600°C in case of non protected loss of active systems for the heat removal
Design and operating feedbacks noticeable (Peach Bottom, AVR, THTR, Fort Saint Vrain, etc.)
Advanced maturity, but for limited reactor powers
INCONVENIENTS and/or INCUMBENT DIFFICULTIES:
Fuel cycle open (but some studies aiming at the closure of the cycle have been performed)
Weak efficiency of the coolant
Significant pressures
High or very high temperatures for the structures (internal structures, etc.): materials to develop, specific risks ?
Risk and consequences of big breaks on primary circuit (mechanical consequences, graphite oxidation or fire, etc. ?)
In service inspection: ?
Risks for the reactor due to industrial linked process
GFR
Gas fast reactors
Power: 600 to 2400 MWth
Coolant: Helium, under pressure
Moderator: none
Fuel: Depleted uranium and plutonium (nitride or carbide)
Coolant temperature at the outlet of the core: 850°C
[www.gedeon.prd.fr]
ADVANTAGES:
Fuel developed for a good resistance at high temperatures (ceramic clad), in case of an accidental loss of heat removal
Very low « void effects » in GFR, vs. SFR
Helium is chemically neutral
A the equilibrium, only the necessity of natural uranium for the fuel re-processing / Possible transmutation of M.A.
Projet RNR-G (CEA)INCONVENIENTS and/or INCUMBENT DIFFICULTIES:
Quite high power density (50 to 100 MW/m3)
Very low thermal inertia of the coolant
Redundant emergency heat removal circuits (3x100%)
The accident of depressurization needs a third barrier under
pressure (P ≈10 bar)
In service inspection: ?
No operating feedbacks
Maturity; to develop…
SFR
Sodium Fast Reactor
Power: ~ 3000 MWth
Coolant: sodium
Moderator: none
Fuel: U238+Pu (oxide, carbide, nitride, or metallic alloy with zirconium)
Coolant temperature at the outlet of the core: 500°C to 550°C, perhaps more
Sodium Fast Reactor
Design and operating feedbacks:
Rapsodie, Phénix, Superphénix, « RNR1500 » project
PFR
JOYO, MONJU
SNR 300
BN 350, BN 600
FBTR (India)
EFR project
…
Loop vs Pool
Coupled vs. Modular
Specific requirements for SFR
Designers need explicit feedbacks on the design and operation of integrated and loop types SFR (Phénix, Superphénix, PFR, Monju, BN 350 and 600, etc.)
Expertise of structures and materials, irradiated or not, from Rapsodie, Phénix, etc. would be of great interest (what is planned for PFR, etc. ?)
Lessons learned from existing probabilistic studies (Superphénix, etc.) would be useful
Specific requirements for SFR
Need for the designers to acquire and account for specific experience upon operating experience of past and
existing reactors such as Superphénix, PFR, Monju, BN 350 et 600, etc.
Perform post operation analysis of irradiated materials from Rapsodie, Phénix, etc. (what is planned for PFR, etc. ?)
Account for and take advantage from existing PSA studies (Superphénix, etc.)
Include the design and technological advances from « RNR 1500 » et EFR design experience
Collect all available information upon FR fleet, worldwide
Identify all progress axis and share the R&D effort
Overview of SFR design and operation safety aspects
based on French experience
MAIN EVENTS WHICH HAVE AFFECTED PHENIX AND
SUPERPHENIX
MAIN SAFETY ASPECTS ADDRESSED CONCERNING
PHENIX AND SUPERPHENIX
Main Events
Sodium-water reactions (PX)
Stability and vibrations of internal structures (SPX)
Assembly plugging (SPX)
Leak of the drum vessel (SPX)
Negative Reactivity Trips (AU/RN) (PX)
Air ingress in the cover gas (SPX)
Argon leak on an intermediate exchanger argon bell (SPX)
Main Safety Aspects 1/4
Sodium void effects
Residual power removal
Severe accidents and initiators
Sodium confinement
Inspectability of structures
What to learn from Phenix End-of-Life Experiments
Main Safety Aspects 2/4
Residual power heat removal
Phénix and Superphénix: residual power underestimated at the design stage
Heat removal by overestimated radioactive transfers
for Superphénix, installation of RUR
for Phénix, de-rating from 600 to 400 MWth
« Total loss of electric supply »: demonstration of a global natural convection is not straightforward
Main Safety Aspects 3/4
Severe accidents and initiators
Void and compaction effects:
As early as the design stage for Phénix and Superphénix, consideration of core meltdown included in authorization decrees
Several aspects:
prevention (PSA required for SPX)
choice of a « symbolic » sequence (SPX) or arbitrary in $/s (PX)
thermal and mechanical energy calculations
(500 MJ for PX and 800 MJ for Superphénix)
Main Safety Aspects 4/4
What to learn from Phenix End-of-life Experiments
Two series of experiments are planned befor the decommissionig of Phenix Reactor :
-Reactor Physic basisc tests such as:
- assembly deplacement,
- introduction of voided zone …
- rod drop,
-Operation test (experimental search of the « equilibrium temperature » in natural convection)
-
ADVANTAGES:
At the equilibrium, fueled with recycled Pu and natural uranium only / Transmutation of M.A. (Am, Cm, Np)
Margin of 200°C / sodium boiling in normal conditions
No significant pressures in the circuits in sodium
Very efficient coolant, large inertia of the reactor in case of loss of convection in the circuits
In case of a first barrier leakage or damaging, the sodium acts like a filter for volatile fission products (iodine, cesium, etc.)
Sodium circuits under low pressures (few bars)
Important and useful design, operating and licensing feedbacks, with Phénix, PFR, Superphénix, etc.: advanced maturity
INCONVENIENTS and/or INCUMBENT DIFFICULTIES:
High power density (300 MW/m3 for Superphénix)
Possibility of positive reactivity injection in case of a sodiumvoiding (boiling, bubble ingress, etc.)
Chemical sodium risks: strong reactions with water, and air (with important consequences in case of spray fires)
Large plutonium inventory
Very hard in service inspection:
Difficulties for an emergency core unloading
Needs for the GEN-IV SFR
Search for cores with low « void effects »
Better assessment of local meltdown propagation risks in the core and re-criticality risks is needed
Sodium: a few risks are still to be identified
Easy access and control must be provided