GENERATION IV NUCLEAR REACTORS

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


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).


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


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


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



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 ?


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


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


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


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)


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


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

Documents

GENERATION IV NUCLEAR REACTORS